NUREG/CR-4652

ORNL/TM-10059

IOAK RIDGE

r.NATIONAL..

t LiABORATORY-'*I

W , ,., .r C, .

-Concrete-Component Aging and

-Its Significance Relative to

Life Extension of Nuclear

Power Plants

D. J. Naus V1

7

1

I

!"".'ERATED BY

i.,MARTIN MARIETTA ENERGY SYSTEMS,

jFOR THE UNITED STATES

DEPARTMENT OF ENERGY..

- - Prepared for the U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Under Interagency Agreement DOE 40-551-75

I : : I 1 2 - .- - I . .

I . . 1. - I I , "'

INC.
NOTICE

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-- Washington, D.C. 20013.7982

and

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NUREG/CR-4652

ORNL/TH-10059

Dist. Category RV

Engineering Technology Division

CONCRETE COMPONENT AGING AND ITS SIGNIFICANCE RELATIVE

TO LIFE EXTENSION OF NUCLEAR POWER PLANTS

D. J. Naus

Manuscript Completed - August 14, 1986

Date Published - September 1986

Prepared for the

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Under Interagency Agreement DOE 40-551-75

NRC FIN No. B0828

Prepared by the

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, Tennessee 37831

operated by

MARTIN MARIETTA ENERGY SYSTEMS, INC.

for the

U.S. DEPARTMENT OF ENERGY

Under Contract No. DE-AC05-840R21400
iii

CONTENTS

Page

LIST OF FIGURES .................. vii

LIST OF TABLES o .o.o.o..oo...o.o.oooooooooosoos !Xi

1. INTRODUCTION .. o............................................... 2

1 .I Background *.............................................. 2

1.2 Objective .... .... .. a ... ........................... ** * * 5

1.3 Approach -....................... 6

References '................................... 6

2. DESCRIPTION OF SAFETY-RELATED CONCRETE COMPONENTS IN LWRs *..... 7

2.1 Design Considerations ............................... 7

2.2 Seismic Category I Structures ..... ....................... 8

2.2.1 Reactor containment buildings ..................... 8

2.2.2 Containment base mats ............................ 21

2.2.3 Biological shield walls and buildings 21

2.2.4 Auxiliary buildings ............................... ' 22

References 27

3. PERFORMANCE OF CONCRETE COMPONENTS IN BOTH NUCLEAR AND

NON-NUCLEAR APPLICATIONS ; '29

3.1 Concrete Longevity ............ ........ 29

3.2 History of the Performance of Concrete Components

in Nuclear-Safety-Related Applications ...... 31

3.2.1 Prestressed concrete containments ................. 33

3.2.2 Prestressed concrete reactor vessels 34

3.2.3 Miscellaneous reactor structures 35

3.3 Problems Experienced with Concrete Material Systems in

Both General Civil Engineering and Nuclear Components .... 37

3.3.1 General civil engineering components 37

3.3.2 'LWR concrete components ........................... 38

3.4 Trending Observations of the Performance of

Concrete Components ....... ........ '.' ... .. . '... *. 44

References - .......C&--''45

4. POTENTIAL ENVIRONMENTAL STRESSORS AND AGING FACTORS TO WHICH

; LWR SAFETY-RELATED CONCRETE COMPONENTS MAY BE SUBJECTED ....... 48

4.1 Concrete Degradation ..-.................... *..I..... ; 0 '.' ' ' ^-.'.... 50

4.1.1 Concrete cracking ............................... 00 51

4.1.2 Aggressive environments ........................... 53
iv

Page

4.1.3 Embedment corrosion * .............................. 55

4.1.4 Extreme environmental exposure .................... 55

4.2 Concrete Reinforcing Steel Degradation ................... 71

4.2.1 Corrosion ........................................ 71

4.2.2 Elevated temperature effects ...................... 73

4.2.3 Irradiation effects ... .............. ............. 73

4.3 Concrete Prestressing Steel Degradation .................. 75

4.3.1 Corrosion ....... ............... .... ... *9**** *....9 75

4.3.2 Elevated temperature effects ..... ................. 76

4.3.3 Irradiation effects .................... ........... 78

4.4. Anchorage Embedment Degradation .......................... 78

References ................................ 79

5. CURRENT TECHNOLOGY FOR DETECTION OF CONCRETE AGING

PHENOMENA ..................................................... 87

5.1 Evaluation of Concrete Materials ......................... 87

5.1.1 Detection of concrete cracking, voids, and

delamination ...................................... 87

5.1.2 In-situ concrete strength determinations .......... 91

5.1.3 Mix composition analysis of hardened-concrete ..... 93

5.2 Evaluation of Mild Steel Reinforcing Materials ........... 94

5.2.1 Visual inspection ........ . . . . . . . . . . . .. ...... . 94

5.2.2 Mechanical and ultrasonic tests ................... 95

5.2.3 Core sampling and chemical and physical tests ..... 95

5.2.4 Potential mapping ............... ......... ... 95

5.2.5 Rate of corrosion probes ..... ..................... 95

5.3 Evaluation of Prestressing Steel Materials ............... 96

5.4 Evaluation of Anchorage Embedments ....................... 97

5.5 Recommended Techniques for Concrete Component

Inspection ................................. 97

References ................................. 99

6. REMEDIAL MEASURES FOR REPAIR OR REPLACEMENT OF DEGRADED

CONCRETE COMPONENTS ........ ... .... .... ........................ 103

6.1. Materials for Repair of Concrete ..... .................... 103

6.1.1 Epoxy resins ................................. 103

6.1.2 Shotcrete ................................. 104

6.1.3 Preplaced aggregate concrete ..... ................. 104

6.1.4 Epoxy ceramic foams ............................... 105

6.1.5 Replacement mortar or concrete .................... 105

6.1.6 Wedge anchors and additional reinforcement 999@9999 105

6.1.7 Miscellaneous sealant materials .' .................. 106

6.2 Preparation of Concrete for Repair * ...................... 106
v

Page

6.3 Techniques for Repair of Concrete ........................ 106

6.3.1 Crack repair methods ............. ................. 107

6.3.2 Spalling or delamination repair methods ........... 108

6.3.3 Nonvisible void repair methods .................... 109

6.3.4 Fractured or shattered concrete repair methods .... 109

6.3.5 Retrofitting (strengthening) methods .............. 109

6.4 Effectiveness of Repairs to Concrete Structural

Components * .............................................. 115

6.4.1 Concrete-rebar bond ............................... 115

6.4.2 Reinforced concrete beams ......................... 116

6.4.3 Concrete joints 0.0eo............................. 117

6.4.4 Concrete walls under fire exposure ................ 119

6.4.5 Earthquake-resistant structural wall .............. 121

References ................................... 122

7. CONSIDERATIONS FOR DEVELOPMENT OF A DAMAGE METHODOLOGY TO

ASSESS DURABILITY FACTOR DETERIORATION RATES AND TO PREDICT

STRUCTURAL RELIABILITY ........ *.************... . .**.o*o**** 125

7.1 Development of Representative Material Property

Data Base ................................................ 126

7.2 Accelerated Aging Methodology for Concrete Materials ..... 126

7.3 Methodology to Provide a Quantitative Measure of

Structural Reliability .0..........0.........00000....... 127

7.3.1 Component classification scheme ................... 131

7.3.2 Methodology for conducting a quantitative assessment of the presence of active deteriorating

influences and their effects ..... ................. 131

7.3.3 Structural reliability technique development for

life extension evaluations ... ..................... 133

References .................................................... 135

8. SUMMARY, CONCLUSIONS, AND RECOMMENDATIONS ........... o.......... 137

8.1 Summary .................................................. 137

8.2 Conclusions .............................................. 138

8.3 Recommendations ........... ............................... 138

Reference .............. oo o...............ooe0.. *oo........ 139

APPENDIX A. LICENSED U.S. POWER REACTORS AS OF APRIL 30, 1985 .... 141

APPENDIX B. ANNOTATED LISTING OF PROBLEM AREAS ASSOCIATED WITH

CONCRETE COMPONENTS IN LWR APPLICATIONS .... 0 0...... 147
vii.

LIST OF FIGURES

Figure Page

1 Average construction time for U.S. nuclear plants . 3

2 LWRs-licensed during .5-year time intervals 4

3 PWR ice condenser containment-configuration -............. 12

4 BWR Mark I containment configuration ..................... 13

5 PWR subatmospheric containment building .................. 14

6..- Indian Point containment building ........................ 15

7 BWR ark II containment configuration ..................... 17

8 BWR Mark'III containment configuration' . 18

9 PWR large dry containment configuration . .................. 19

10 PWR hemispherical dome.- three buttress containment ...... 20

11 Reinforcements in Indian Pointcontainment cylinder ...... 22

12 Reinforcements and tendons in Zion containment cylinder es 23

13 Section through Grand Gulf containment foundation mat ... 24

14 Generic nuclear power plant building layouts ............. 25

15 - Typical Category I structure ............................. 26

16 Compressive'strength--ge relations for SeriesA and B

.. concrete-stored indoors and.outdoors ...................... 32

17 C reactor cracks in biological shield at SRP .............. 36

18 Distribution of LWR concrete component problem areas '.... 39

19 Extent of dome delamination'for Turkey Point Unit 3 ..... 40

'20 Extent of dome delamination for Crystal River Unit 3 ..... 41

21A Factors in production of good concrete; ................... 45

22 Effect of temperature exposure -on compressive strength

of concrete hot testing. .'...**... .. *eos.00000000000*0 57

23 Effect of temperature'exposure on compressive strength -

-of concrete cold testing 58

24 Effect of temperature exposure on modulus of- elasticity

of concrete hot and cold testing *....... ............ 59

25 Stress-strain diagrams of sealed limestone concrete

s p ecimens o... ' ' o9 oa 9 ' ' a ' ' ao59

26 Stress-strain diagrams of unsealed limestone concrete

specire of cocrte w different aggrgats ....... ' ' ' 60

27 Creep of concretes- with different aggregates ........................ 60
viii

Figure Page

28 Creep of portland cement/porphyry concrete at various

temperatures 61

29 Effects of temperature cycles on limestone concretes ..... 62

30 Laboratory vs actual sample compressive 'strength data we** 64

31 Thermal neutron distribution in ordinary concrete as a

function of temperature 65

32 Compressive strength of concrete exposed to neutron

radiation relative to untreated concrete: thermal

effects on strength (a) not included, (b) included ....... 68

33 Tensile strength of concrete exposed to neutron

radiation relative to untreated concrete: thermal

effects on strength (a) not included, (b) included ....... 69

34 Effects of fast neutron exposure-on volume change of

flint aggregate and limestone aggregate concretes ******a* 70

35 Modulus of elasticity-of concrete exposed to neutron

radiation.relative to untreated concrete: thermal

effects on modulus not included ... 70

36 Cross section of ORNL graphite reactor shield ............ 71

37 Effect of temperature on yield strength (0.2%) of four

types-of German reinforcing steel .... .73

38 Stress-strain curves for unirradiated and irradiated

mild steel 74

39 Charpy V-notch energy/temperature-curves for

unirradiated and irradiated mild steel 74

40 Effect of temperature on ultimatestrength of 30

different types of prestressing steel 76

41 Stress-strain curves for ANSI/ASTM A 421 steel at

various temperatures, 77

42 Relaxation-vs time curve for low-relaxation strand

stressed to 70% GUTS and held at various temperatures *see 77

43 Conditions that influence transmission of sound in

concrete 89

44 Typial Typical shear wall strengthening 110

45 Typical methods for column strengthening 111

46 Typical methods for strengthening beams and new

collector members 112

47 Typical ,fountTyp icalugmentation augmentation 113

48 Typical anchoragefor parapet anpets and cornices 114

49 Details of pull-out and reinforced concrete beam

test specimens s**.....115
ix

Figure Page

50 Dimensions, loadings, and reinforcement details of reinforced concrete beams containing a large rectangular

opening .......................... 116

51 Schematics of tension and shear crack-type reinforced

concrete beam test specimens 118

52 Details of concrete pushoff test specimens 119

53 General configuration of small-scale expoxy-repaired

shear wall specimens .................. 120

54 Nominal dimensions of one-third-scale five-story

wall test specimens o.... . . ...... 121

55 Factors influencing the successful repair of a

concrete component ................ 125

56 ASTM E 632 recommended practice for developing

predictive service life tests ............................ 128

57 Systematic evaluation method for determination of

structural durability factors ....................... 132

58 LWR concrete component evaluation methodology 134
xi

LIST OF TABLES

Table Page

1 Summary of containments for U.S. power reactors ........... 10

2 Summary of design parameters for selected containment

systems 11

3 Interacting factors for mechanisms producing premature

concrete deterioration 49

4 Predominant environmental stressors to which safetyrelated components in a LWR may be subjected ........ *...... 50

5 Causes of concrete cracking ............................... 51

6 Recommended nondestructive evaluation methods for

inspection of concrete materials ........................ 98

7 Several mathematical models used in accelerated aging

studies 129

8 Major steps and schematic diagrams for analysis of data

obtained from a prototype accelerated life test o........ 130
7CONCRETE-COMPONENT AGING AND ITS SIGNIFICANCE RELATIVE

TO LIFE EXTENSION OF NUCLEAR POWER PLANTS* -

D. J. Naus

ABSTRACT -

The objectives 'of this study are to (1) expand upon the

work that was initiated'in 'the first 'two Electric Power Res'earch

Institute studies relaTtive td-6lbngevity and 'life extension considerations' of safety-related ;concrete' components' in 'light-' 

water 'reactor (LWR) facilities 'and (2)'provide background 'that

will logically lead to'subsequent development:of a methodology

for assessing and predicting the effects of aging on the 'per-'

formance of'concrete-based materials nd components.' These objectives'are consistent with Nuclear Plant Aging Research (NPAR)

Program goals't (1) to-'identify and characterize aging and service wear effects that,' if unchecked,'icould cause degradation of

structures, components, and systems'and," thereby, impair plant

safety; (2) to identify methods of inspection, surveillance,

and monitoring'or of evaluating residual life of'structures,

components, and systems that will ensure timely detection of.

significant'aging effects before loss of-safety function; and-'

(3) to 'evaluate -the effectiveness of 'storage' maintenance, repair, and replacement practices in mitigating-the rate and extent'of ,degradation caused by ging'and service wear.

Applications of safety-related concrete 'components to LWR

technology 'are 'identified, and pertinent-components (contain- -

ment buildings,'containment base mats, biological shield walls

and'buildings,- and auxiliary buildings),'as well as the materials of which they are constructed (concrete, 'mild steel'reinforcement,' prestressing'system's, embedments, and anchorages)

are described. Historical-performance of concrete components

was established through information presented on concrete longevity, component behavior in both LWR and high-temperature

gas-cooled reactor applications, and areview of problems with

concrete components in both general civil engineering and nuclear power applications. The majority of the problems identified in conjunction with:znuclear' power applications were minor

and involved either concrete cracking, concrete voids, or low

-' concrete-strengths at early!.ages.- Five incidences involving .

LWR concretecontainments considered significant are described

'*Research 'fundea by the Nuclear Regulatory Commission NPAR Program

and conducted in conformance with its program goals.

B. M. Morris'and J. P.'Vo'ra,1NucZear PZant Aging 'Research (NPAR)

Program Plan, NUREG-1144, Division of Engineering' Technology, .'Office -of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington, D.C., July 1985.
2

in detail from occurrence and detection through remedial measures used to restore structural integrity or continuity.

These incidences were related either to. design, construction,

or human error and involved two dome delaminations, voids under

tendon-bearing plate, anchor head failures, and a breakdown in

quality control and construction management.

Potential environmental stressors and aging factors to

which LWR safety-related components could be subjected are

identified and discussed in terms of durability factors related

to the materials used. to fabricate the components (e.g., concrete, mild steel; reinforcement, prestressing systems, and embedments). -The current technology for detection of concrete

aging phenomena is. also presented in terms of methods applicable to the particular. material system that could experience deteriorating effects. Remedial measures for the repair or replacement of degraded.concrete components are discussed, and

examples of prerepair and postrepair structural performance are

presented to indicate the effectiveness of these measures.

--Finally, considerations relative to development of a damage

methodology for assessment of durability factor deterioration

rates and prediction of structural reliability are discussed.

Conclusions and recommendations of the report note the

need for (1) obtaining aging data from decommissioned plants,

(2) using in-service inspection programs to provide aging

trends, (3) developing a methodology to quantitatively and uniformly assess structural reliability as affected by aging or

degradation of.structural materials, and (4). performing research in support of all these needs. Although, as a group

concrete structures have a history of reliability and durability, there is no standardized, widely accepted methodology

for quantifying the condition and capacity of an individual

structure. Such a means of, evaluation needs to b developed if

informed licensing decisions are to be made on an extension of

licensed design life of nuclear power plant structures.

1. INTRODUCTION

1.1 Background

Nuclear power currently supplies 16% of the US. electricity requirements. This percentage is expected to rise to -20%lby 1990.*

Despite the increasing role of nuclear energy in power production, the

current trend is toward-completion (or cancellation) of plants under

*As of August-1985, there were 95 licensed U.S. power reactors.

Appendix A presents a listing.
3

construction, with-no new nuclear plants having been ordered since the

1970s.* ,

Although the cessation of orders for nuclear power plants has resulted in a large degree from a slowdown of the growth in demand for

electricity,'a number of other factors have eroded the economic-advantage

-nuclear power once-had over many other forms of energy production. Escalation of material-and labor costs, higher interest rates, etc., havelresulted in a'significant increase in-the average duration of plant construction (see Fig. 1) and -almost an order of magnitude increase in cost

-of generating capacity additions since the mid-1960s.1 These factors -

have resulted'in'hesitancy on the part of utilities to-consider the construction of new nuclear power plant~facilities. 'In addition'two -other

factors ust be considered relative to the ability of the utility industry to r-Fet the future energyrequirements: design lifetime and shutdown

costs of existing nuclear plants.

The basic laws that regulate the'-design (and-construction) of nuclear plants are contained-in Title 10 of the Code of Federal Regulations

*As of August 1985, 30 plants 121 pressurized-water reactors (PWR),

9 boiling-water reactors (BWR)] were under construction, 90,plants

(60 Ps, 30 BWRs) were canceled or indefinitely deferred, and,2 plants

(PWRs) are-planned. 

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:NUMBER OF PLANT 7; - (79

-- - 6 -i 1

(46) '56) >1NUMBER OF MONTHS - , -

12 .4 -- - -~~~~~~~~~~~~12 ,

> *88) - -~ - -- - * 1982 DOE

- - - ~~~~~~~~~~* 1985A1F - - ; - . ~~~~~~~~PROJECTEDI~~~~~~~~~~~~~(0 -''I 'I' ' (11'i 109) '-1 ' I'

25 BEFORE

70 70 71 72 73 74 75 76 77 78 79 80 81 82

YEAR OF COMPLETION , I

83 84 85 86 87 88 89

Fig. 1. Average construction time for U.S. nuclear plants.
4

(10 CFR),2 which is clarified by Regulatory Guides, NUREG reports, Standard Review Plans, etc. The design lifetime of nuclear plants is somewhat.unique because the operating license for a plant. has a specific expiration date,.usually 40 years from the date of the construction permit

issuance.. Figure 2, which presents a histogram of light-water reactor.

(LWR) plants listedin Appendix A as a function of years since an operating license was-granted, indicates that plants will start to reach termination of their operating licenses in the next 15 to-20 years.- The potential impact.of the expiration of operating permits.is-further clarified

in Ref..3, where it is noted, that under the present situation* the United

States could experience a loss of electric generating capacity on the

order of 150 GWT during the time period 2005 to 2020.

*Assumes no life extension of facilities.

tA more recent estimate of the potential loss-of electric generating

capacity indicates that the loss is on the order of 50 to 60 GW.

YEARS SINCE OPERATING LICENSE

ORNL-DWG 86-4553 ETD

25-

29

20- 15- 10-

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1961

to

1965

1966

to

1970

1971

to

1975

1976

to

1980

1981

to

1985

YEAR OPERATING LICENSE GRANTED

Fig. 2. LWRs licensed'during 5-year time intervals.
5

Final shutdown and decommissioning-costs are other important considerations of the utilities. As noted in Ref. 4, estimated dismantlement

costs in terms of 1983dollars range from'$14.8 Tillion for Prairie"'

Island 2 to $333'million for Calvert Cliffs 1 Although'these costs are

small-whencompared with the initial and-lifetime costs-of'a'plant, they -

must be built into'the rate structure based on an'agreement between-the

principal 'ftili'ty owner and the state regulatory commission. -

A potential timely and cost-effective solution tthe problem of

meeting future-energy demand is to-extend'the service life of the nclearplants. Refurbishment and life extension activities have worked well-for

non-nuclear generating plants, with some fossil-fueled plants'having been

in service'for 50 to 60 years. Hydroelectric plants are expected to--operate-for significantly more than-40 years3. Refurbishment and life

extension should work equally.well for nuclear plants,'especially because'

many of the plants may have only been in operation 25 to 30 years prior

to epiration-of their operating'license.

Two' Electric Power.Research Institute (EPRI) studies3 V5 have already

addressed this-topic.' In the first study3 theXconcept of operating a LWR

power plant beyond its initial license-life was studied'from both economic

and technological aspects. Conclusions of the study were that ()-extending the life of LWRs'was economically beneficial if,' up'to the first

decadeof the'21st century,-the nuclear'fuel'costs remainilow relative to

other feasible basel6ad power generation-technologies (allowable-downtime

for refurbishment can 'be several years and cost 'onthe order of several

100 million 1979 dollars before'economic feasibility becomes borderline);

(2) even in situations where a large piece of equipment such as a reactor

pressure vessel or steam generator required replacement, case studies

have shown that replacement is feasible;- and (3) major repairs or replacement of the concrete structure in the base mat, containment walls,

or biological shield'within the containment could result in:a significant

cost. The' second EPRI-funded study5 considered'the feasibility.'of extending the,-life of existing nuclear'power plants and-concluded that

power reactors should have useful service lives substantially in excess

of the licensed 40 years from the date of construction and that a generic

method for verifying 'the continued integrity of concrete structures

should be developed.

1.2 Obective 

'The objective of this study is'twofold: (1) to' expand upon' the' work

that was initiated in the first two EPRI studies3'5 relative to' the'longevity and life extension considerations of safety-related concrete components in LWR facilities 'and (2)'to provide a backgr6und that will logi-'

cally lead to subseqtuent development 'of a methodology forassessing and

predicting the effects of aging* on the performance of the concrete-based

materials and components.. . .

*For definition of aging refer to NUREG-1144.6
6

1.3 Approach

Information in the EPRI studies3 v5 indicated that concrete durability (aging) under the influence of either material interactions, aggressive.environments.(freeze-thaw, wetting-drying, or chemical), or exposure to extreme environments (elevated temperature, irradiation, or

seismic) is one of the-key issues in nuclear plant life extension. Al-.

though operating plants have reported little difficulty with concrete

materials, an evaluation of the long-term effects of the environmental

challenges to which these structures are subjected has not been adequately addressed. : - -

.The approach to be followed in accordance with the Nuclear Plant

Aging Research (NPAR) strategy to evaluate the long-term environmental

challenges of LWR concrete facilities and thus provide the background

material to meet the previously stated objectives will consist of six

parts: (1) description of primary safety-related concrete components in

LWRs; (2) review of the performance of concrete components in both nuclear and non-nuclear applications; (3) identification and discussion of

potential environmental stressors and aging factors to which concrete

safety-related components may-be subjected in an LWR environment; (4) review of the current tate of the art for in-service-inspection, surveillance, and detection of concrete aging phenomena and assessment of structural adequacy; (5) discussion of remedial measures for the repair or

replacement of degraded concrete components; and (6) remarks concerning

correlations between damage assessment and life extension considerations.

References

1. R. J. Christensen, "An Architect-Engineer Perspective," pp. 25-40 in

Proceedings of a Conference on Construction of Power Generation

Facilities, J. Willenbrock, ed., The Pennsylvania State University,

University Park, Pa., September 16-18, 1981.

2. 10 CFR.

3. C. A. Negin et al., Extended Life Operation of Light Water Reactors:

Economic and Technological Review, EPRI NP-2418, vols. 1 and 2, Electric Power Research Institute, Palo Alto, Calif., June 1982.

4. "Public Interest Effort Surveys Decommissioning Tactics and Funding,"

Nucleonics Week 26(17) (April 25, 1985).

5. I., Spiewak and R. S. Livingston, The Longevity of Nuclear Power Systems, EPRI NP-4208,. Electric Power Research Institute, Palo Alto,

Calif., August 1985.

6. B. M. Morris and J. P. Vora, Nuclear Plant Aging Research (NPAR) Program Plan, NUREG-1144, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington, D.C., July 1985.
7

2. DESCRIPTION OF SAFETY-RELATED CONCRETE

COMPONENTS IN LWRs

2.1 Design Considerations

General Design Criteria 1, "Quality.Standards and Records"; 2,."Design Bases for Protection Against Natural Phenomena"; and 4, "Environmental and Missile Design Bases," *of Appendix A,."General Design Criteria

for Nuclear Plants," to 10 CFR 50, Licensing of-Production and Utilization Facilities,," require, in part, that structures, systems, and components important to safety be.designed, fabricated, erected, andItested to

quality standards commensurate with the safety functions to be performed

and that they be designed to withstand the effects of postulated accidents and environmental conditions associated with normal operating conditions.

Initially, existing building codes such as American Concrete Institute (ACI);Standard 318-71, BuiZding Code Requirements for Reinforced

Concrete (ANSI A8Q .1-1972),2 were used in the nuclear industry as the

basis for the design of'concrete structural-members. However, because

the existing building codes did not cover the entire spectrum of.design

requirements and because they were not always considered adequate, the

Nuclear Regulatory Commission (NRC) developed its own criteria for the

design'of Category.I structures.*- In particular, definitions of load'.

combinations for'both oerating and accident conditions were provided, as

well as'a list of tornado-borne'missiles and a descriptionof the characteristics of tornados for different regions of the United States."

Using-ACI 318-71 as a basis, with modifications to accommodate the

unique performance requirements of nuclear plants, ACI Committee 349 developed and published in October 1976 ACI 349-76, Code Requirements for

Nuclear Safety Related Structures.3 The procedures and requirements described in this document are generally acceptable to the NRC staff and

provide an adequate basis for complying-with the general design criteria

for structures other than.reactor vessels and containments.t Conditions

for applying the procedures and requirements in ACI 349 are presented in

Ref; 4, and-additional information-on the design of seismic Category I

structures,''which-are required to remain functional.if'the Safe Shutdown

Earthquake (SSE) occurs, are contained in Ref. 5. *Reference,6,presents a

good comparison between ACI 318' and ACI 349.

Requirements for the design of concrete reactor vessels and containments are presented in ACI 359-77, ASME Section' III- Division 2, Code

*Category-I structures are those essential to the function of'the

safety class systems'and components, or that-house, support,'-or protect

safety class systems or components,;.and whose failure could lead to loss

of function of-the safety class system and components housed, supported,

or protected. -

tACI 349-76 is endorsed by U.S. Nuclear Regulatory-Guide 1.142,

"Safety-Related"Concrete Structures for Nuclear Power Plants (Other than

Reactor Vessels and Containments)."4 :
8

for Concrete Reactor Vessel and Containments.* 7 Supplemental load combination criteria are presented in Sect. 3.8.1 of the NRC Standard Review

Plan NUREG-0800].8

2.2 Seismic Category I Structures

A myriad of concrete-based structures are contained as a part of an

LWR system. Although the particular components may vary somewhat according to the selection of nuclear steam supply system (NSSS) and containment concept, the structures can be grouped into four primary categories

for discussion: reactor containment buildings, containment base mats,

biological shield walls and buildings, and auxiliary buildings (balanceof-plant structures).

2.2.1 Reactor containment buildings

2.2.1.1 Background. From a safety standpoint the containment

building is probably the most'important structure-of a nuclear power

plant facility because it serves as the final barrier against the release

of radioactive fission products to the environment under postulated design'basis accident (DBA)'conditions. Containment design is based on

pressure and'temperature loadings associated with a loss-of-coolant accident'(LOCA), resulting'from a double-ended rupture of the largest size

pipe in the reactor coolant system. The containment is also designed to

retain its integrity under low probability (<10-4) environmental loadings

such as those generated by earthquake, tornado, or other site-specific

environmental events such as floods, seiche, or tsunami. Additionally,

the: containment 'is' required to'provide biological shielding under both

normal'and' accident conditions and is required to protect the internal

equipment from external missiles, such as tornado- or turbine-generated

missiles and aircraft impact (where postulated). Design pressures and

temperatures 'are dependent on containment-free volume and presence of

either'heat sinks or pressure, suppression systems.

.'2.2.1.2 Evolution. Prior to 1965, installed capacity of nuclear

power plants in the 50- to 400-MW(e) range utilized steel containments of

various configurations, for example, spherical, cylindrical with elliptical ottom and hemispherical'top, and cylindrical with hemispherical dome

and flat slab. Their designs conformed to the American Society of

Mechanical Engineers (ASME) Unfired Pressure Vessel Code,10 with the

siells fabricated'from welded steel plates up to 38 mm in'thickness.

Support for the reactor vessel and shielding requirements is provided by

reinforced concrete. As the plant sizes were increased to 800-MW(e),

shielding requirements increased, and the practical limit for fabrication

of steel containments without requiring postweld heat treatment were exceeded. At this time it also seemed prudent to combine the containment

and shielding functions into a composite steel-lined reinforced concrete

structure.

- *ACI 359-77 is endorsed by U.S. Nuclear Regulatory uide 1.136,

"Material for Concrete Containments."9
9 

The first concrete containments were built in the mid-1960s and

typically consisted of an 1.4-m-thick-cylindrical reinforced concrete

wall with an 1.1-m-thick hemispherical dome and a flat base slab. Leak

tightness was provided by a steel'liier, which generally ranged from 6.35

to 12.7 mm in thickness depending oji its location. Grade 60 (Nos. 11,

14, and 18) reinforcing barsfwere normally utilized to resist hoop,

axial, seismic, and shear loadings. Concrete compressive strengths

ranged from 20.7 to 34.5 MPi*. 'Later "the concrete was partially prestressed in the vertical direction only with mechanically spliced reinforcing steel in'the hoop'direction and dome (Ginna, Robinson 2).

Fully prestressed containments were first built in the late 1960s.

The first generation of fully prestressed containments typically consisted of an 1.1-m-thick cylindrical wall, an 0.8-m-thick shallow (ellipsoidal) dome, a large ring girder at the intersection of.the dome and

wall, six buttresses, and'a flat reinforced'concrete base slab. The wall

was prestressed by hoop tendons, anchored at'two buttresses'and spanning

1200, and vertical tendons placed with equal'spacings near the inside and

outside faces. Dome prestressing was provided by three groups of tendons,

with each group at 1200 with respect to the other two groups. Tendons

consisted of ninety 6.35-mm-diam steel wires posttensioned to 1.65 GPa.

Grade 40 or Grade 60 rebars (Nos. 9, 10, 11, and 18) were used to provide

light reinforcement. Concrete strengths ranged from 27.6 to 34.5 MPa.

As a consequence of the quantity of tendons (>900), which required a

very labor-intensive activity to fabricate, position, tension, and corrosion proof and-an increase'in plant'size, the 'second generation of fully

prestressed concrete containments was developed. In the second generation containments the number of buttresses was reduced to three, with the

hoop tendons spanning 240. between buttresses.' This resulted in both a

reduction in installation time-and an improvement in the radial force

distribution on the shell.1 1 Another change was that the capacity of'the

prestressing tendons was approximately doubled,'which was reflected in.a

reduction of up to 510 tendons- (depending on containment height, diameter, and design pressure) relative to first generation design requirements.11

. Tendon systems utilized were composed of either one hundred

eighty 6.35-mm-diam steel-wires or fifty-five 12.7-mm-diam seven-wire

strands. Concrete and reinforcing steel requirements. and the necessity

for aring girder'remained essentially unchanged from the first'generation of prestressed concrete containments.

Third-generation prestressed concrete containments replaced the ellipsoidal dome with'a hemispherical dome, thus permitting'a'simpler'posttensioning tendon layout. Through-the use of inverted U-shaped vertical'

tendons, the ring "girder was eliminated. The inverted U-shaped tendons-'

were divided into two tendon sets 'oriented 90° to each other'in the dome

array, with all tendons in each set parallel to each other in' the dome

and the midtendon in'each set'located in a diametrical plane of the dome

and cylinder. Hoop tendons were still anchored in a three-buttress ar--

rangement with each tendon spanning-2400. Concrete, reinforcing steel,

and tendon systems were essentially the same as for the second-generation

containments.

'2.2.1.3 Summary description'of containment types'utilized for LWRs

in'the United States.' As of April 30, 1985, there were 95 licensed U.S.,

nuclear power reactors.12 Table 1 presents a summary distribution of'
10.

Table 1. Summary of containments for

U.S. power reactors

Containment Reactor . Containment description

construction 'type Tu . . ~Type -Number

Steel PWR Dry 9

PWR Ice condenser 5

BWR MKI 21

BWR MKII 1

BWR pre-MK 4

Reinforced PWR Subatmospheric 6

concrete PWR Dry 6

PWR Ice condenser 2

BWR, MKI 2

BWR MKII 3

BWR MKIII I

Prestressed HTGR 1

concrete. PWR Dry 32

BWR MKII 2

containment concepts that have been used-for both the PWR and BWR systems.* As noted in the table, a variety of containment designs have been

utilized. Rather than present--a summary description for each design,

only representative PWR and .BWR containment designs for each major type

of containment construction have been -selected to identify major components and indicate design parameters.- Containment designs'selected include: steel (PWR'ice condenser and BWR MKI),t reinforced concrete (PWR

subatmospheric, PWR dry,.BWR MKIIi and BWR MKIII), and prestressed concrete (PWR shallow dome - six buttresses and PWR hemispherical dome -

three buttresses). Table 2 presents representative design parameters for

each of these containment types for which a reference plant has been

selected as an example.

* PWR ice condenser. The containment for each of the Sequoyah reactors consists of a free-standing steel containment vessel,- ice condenser, internal reinforced concrete structure, and reinforced concrete

shield building.13 Figure 3 presents the containment configuration. The

shield-building is a reinforced concrete cylinder 0.9 m thick, with a

2.7-m-thick slab on rock and-a 0.6-m-thick dome. Concrete strengths are

.*See Appendix-'A for a more detailed description'for each reactor

plant. .

---tSteel containment designs have been included to define and indicate

positioning of concrete-components associated with the-containment configuration.
Table 2. Summary of design.parameters for selected containment systems

LWR - - Reference Allowable Containmenta Pressure Temperature

reactor - Containment type plant leak rate free volume (kPa) (C)

type (Doc. No.) (vol /d) (10)3 3) ccident Design Accident Design Accdn ein Acdn

PWR ice condenser Steel cylinder - Sequoyah 0.25 34.0 Atmospheric 82.7 15.6-48.9 104.4

hemispherical' dome (50-327)'

BWR MXKI Steel - pressure Peach Bottom 0.50 4.5 (D) Atmospheric 427.5 (57.2 138.3

suppression (50-277) 3.4 (PSC)

3.9 (PCP)

PWR subatmospheric Reinforced concrete - Storry 0.10 51.0 62.1-75.8 310.3 15.6-40.6 65.6

hemispherical dome (50-280)

PWR dry Reinforced *concrete - Indtan Point, 3 0.10 73.9 Atmospheric 324.1 (48.9

lhemispherical dome (50-286) '

BWR MKII Reinforced concrete - Limerick 0.50 7.1 (D) Atmospheric 379.2 (57.2 171.1 (D)

pressure suppression (50-352) 4.2-4.6 (PSC) 104.4 (PSC)

3.3-3.6 (PCP)

BWR MIII Reinforced concrete - Grand Gulf 0.40 39.6' Atmospheric 103.4 C57.2 85.0

pressure suppression (50-416) 7.6 (D)' - 206.8 (D) 165.6 (D)

- . ! 3.9 (PCP)

PWR shallow dome Prestressed concrete - Zion 0.10 73.6, Atmospheric 324.1 , (54.4 132.8

six buttresses (50-295) , ,

PWR hemispherical. Prestressed concrete - Trojan, 0.20 63.0 Atmospheric 413.7 (48.9 138.3

dome three buttresses (50-344)

aD - drywell

PSC - pressure-suppression chamber

PCP - pressure-suppression chamber pool water.
12

ORNL-DWG 86-4554 ETD

I IL.

I

PRESSURIZER-- BU - t

RELIEF TANK *~ .

NSTRUMENTATION WELL -; ,,

RECIRCULATION SUMP CONTAINMENT PIT

AND LOWER COMPARTMENT w * -- AND........................... *

REACTOR CAVITY

PICKUP SUMP , , ,,..;; .,- ; SM

DOTTED AREAS INDICATE RECIRCULATION SUMP VOLUME

Fig. 3. PWR ice condenser containment configuration. Source: Containment Performance Working Group Report, Draft Report for Comment,

NUREG-1037, U.S. Nuclear Regulatory Commission, Washington, D.C., May

1985.
13

27.6 MPa* for the cylinder dome and walls and 20.7 MPa 'for the base 'slab.

A 5.2-m-ID circular reinforced concrete wall 2.59 m thick (primarily for

biological shielding) suports and encloses the 6.23-MN reactor-vessel

above the lower reactor cavity.' Other-'concrete components include a

variety of walls, divider'barriers, floor -slabs, and columns.-

BWR Mark I. The containment.'at'Peach Bottom and Browns Ferry14'15

is a pressure-suppression system that consists of a drywell,'-a

pressure-suppression chamber (torus) that stores a large volume'of water,

and a connecting vent system between the drywell and water pool (Fig. 4).

The drywell is a lightbulb-shaped steel pressure vessel with a spherical

lower portion and a cylindrical upper portion. The suppression chamber

is a steel pressure vessel in the shape of a torus, which is located

below and encircles the drywell. Eight circular vent pipes connect the

suppression chamber with the drywell. The drywell is enclosed in a reinforced concrete structure for shielding purposes. In areas where it

backs up the drywell shell, the reinforced concrete provides additional

resistance to deformation of the containment shell. Shielding over the

*Concrete strengths are presented as 28-d design'values. Actual

strength levels of the concrete in the structures in all likelihood significantly exceeded these values.

OR NL-DWG 86-4555 ETD

.1MEACTOR BUILDING 5

1111

I

''Fig. 4.' BWR Mark I containment configuration.' 'Source: Containment

Performance Working Group Report, Draft Report for Comment, NUREG-1037,'

U.S. Nuclear Regulatory Commission, Washington,'D.C., May 1985.
14

top of the drywell is provided by removable, segmented, reinforced concrete plugs.,

PWR subatmospheric. The containment vessel at Surry14'16 is a

steel-lined reinforced concrete'structure with an -1.4-mr-thick vertical

cylindrical wall, an O.8-m-thick hemispherical dome, and an 3-m-thick

flat base slab'(Fig. 5). The steel liner for the wail is 9.5 mm thick.

Over the base mat the steel iner consists' of 6.35- and 19.1-mm plates.

Approximately 0.6 m of concrete is placed on top of-the mat liner to protect it from thermal loadings'and internal missiles. The steel liner for

ORNL-DWG 86-4556 ETD

a

REACTOR CAViTY

Fig. 5. PWR-subatmospheric containment building. Source: Containment Performance Working Group;Report, Draft Report for Comment, NUREG1037, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1985.
15

the dome is 12.7 mm thick.- -The containment is fabricated using 20.7- and

27.6-MPa concretes.. Primary reinforcement is.,provided by bars placed

circumferentially and axially in the wall. Seismic reinforcement consists of bars placed helically at an angle of 450 with the horizontal in

both directions in the wall of the cylinder. Stirrups or diagonal bars

are provided in the lower portion of the cylinder ,to resist radial shear.

Internally, reinforced concrete is used for biological shielding, reactor

vessel support, barriers, floors, and, walls. -

PWR dry... The containment building at Indian Point 314,17 (Fig. 6)

is a reinforced concrete structure consisting of an 2.7-m-thick base

mat, an -1.4-m-thick cylindrical wall, and an 1.1-m-thick hemispherical

dome. Concrete strengths.are on the order of 21 to 28 MPa. Leak tightness is provided by a ductile steel liner whose thickness is 6.35 mm over

ORNL-DWG 86-4557 ETD

DOME

0.5 in. Aft, EL 262' - LINER PLATE 3 f t i.

/1 - ; . ~R-67 ft \'

-SPRINGLINE EL 191'

iJ-~~. - EL 43'-

_ . < ~~135 ft I D p _ 

_~~~ t 6 in. -* - .ft 6in.

:0.375 in.' - LINER PLATE:..

. . . . . ; -C~~Y L I NDE R , ~~0.25 in . -; ' 

- -. . . ' -- LINER PLATEFU 

l _ me s _~~~~~ ELE 43'

BASEMAT

- , - 9ft EL 13'

Fig. 6. Indian Point containmentrbuilding. Source: S. Sharma,

Y. K. Wang, and M. Reich, -Vltimate Pressure'Capacity of Reinforced and

Prestressed Concrete Containrents, NUREG/CR-4149, BNL-NUREG-57857, Brookhaven National Laboratory, Upton, New York, May 1985.
16

the base mat, 12.7 mm in the dome and bottom 9.1-m section of the cylinder,'and 9.5-mm over the remaining'height of the cylinder. Containment

reinforcement'consists primarily of Nos. 11, 14, and 18 Grade 60 reinforcing bars. Membrane reinforcement in the cylinder wall and dome is

divided into two groupsplaced near the inside and outside faces of the

containment wall. Each group consists of two layers of hoop bars and one

layer' of meridional'bars. 'A layer of helical bars at +45° with the vertical axis is placed near the outside wall face to resist in-plane seis--

mic forces. Secondary meridional reinforcement is also provided at the

base mat-cylinder intersection to help resist the high bending moments

and shear''forces that -could develop.

BWR Mark II. A pressure-suppression system consisting of a drywell

and suppression chamber separated by a horizontal diaphragm slab is used

as the containment at Limerick.14 The containment is in the form of a

truncated cone over a cylindrical section positioned on a flat foundation

mat (Fig. 7). The upper conical section, which contains the reactor and

high-energy piping systems, forms the suppression chamber. A floor separates the drywell from the suppression chamber, with downcomers providing

communication. The containment is a reinforced concrete structure lined

with welded steel plate and has a steel domed closure head at the top of

the drywell. Reinforced concrete is also utilized internally in the containment as a sacrificial shield wall, reactor support, columns, and

floors.

BWR Mark III. The containment at Grand Gulf14 ,18 ,19 is a pressuresuppression system with the drywell completely enclosed by the containment structure (Fig. 8). The lower portion of the-structure also serves

to form the pressure-suppression pool. The containment is constructed of

34-MPa cast-in-place reinforced concrete and consists of an -1.1-m-thick

right circular cylinder capped by an 0.8-m-thick hemispherical dome and

founded on an -2.9-m-thick flat circular base mat. The inside surface

of the containment is covered by a 6.35-mm-thick steel-liner, which forms

a leak-tight barrier (type,304 stainless steel material is used in the

area below the suppression pool level). Main reinforcement in the wall

consists of inside and'outside layers of hoop reinforcement (No. 18

bars), outside vertical reinforcement (Nos. 10 and 18.bars), and diagonal

reinforcement placed in two directions to form a helix with an angle of

-45° from the vertical axis of the wall (Nos. 11, 14, and 18 bars).

Additional reinforcement is also placed near the intersection of the wall

and foundation'mat' nd areas around major penetrations, pipe penetrations, floor brackets, polar crane brackets, etc. Main reinforcement in

the dome consists of (1) hoop reinforcement composed of inner and outer

layers of circumferential steel bars (Nos. 10, 14, and 18) extending from

the intersection of the dome and cylindrical wall to 46° above the

spring line; (2) inside and outside groups of U-shaped reinforcement

(Nos. 10, 14, and 18-bars) composed of-two mutually perpendicular layers

of steel bars; and (3) diagonal reinforcement (No. 11 bars) continuous

with the diagonal cylindrical wall-reinforcement and extending up to 40°

above the spring line. Additional dome reinforcement includes meridional

bars (Nos. 14 and 18).as a continuation of inside face vertical reinforcement in the cylinder wall. Internal reinforced concrete structures

include reactor support pedestal,'shield walls, drywell walls, weir

walls, etc.
17 '

ORNL-DWG 86-4558 ETD

I

. ; II tI-.

Fig. 7. BWR Mark IV:-containment 'cnfiguration Source: Containment Performance Working droup Report' Draft Report fr Comment, NUREG1037, U.S. Nuclear Regulatory Commission, Washington, D.C', May'-1985.-

PWR shallow dome - six buttresses.- The containment at Zion14'20 is

a steel-lined prestressed concrete structure -with an-1.1-m-thick vertical cylinder-walltand an -0;8-m-thick shallow (ellipsoidal) domed roof

supported on a reinforced concrete-foundation slab .(Fig. 9).'.The containment is fabricated from-f34."5-MPa reinforced concrete, and a6.35-mmthick steel liner is utilized-to provide leak tightness. A large ringgirder is positioned at theintersection of the dome and-wall for-anchorage of the'dome and vertical' prestressing tendons, and six buttresses are
18

ORNL-DWG 86-4559 ETD

D ;ei 2.5-ft CONCRETE

(Detail drawing not to scale)

UPPER CONTAINMENT POOL NOTE: UPPER CONTAINMENT

. =. W r \ ,, _ : ~~~~~DOME AND LOWER

WETWELL COMMUNICATE

WITH EACH OTHER

: rt _ K _TEMPORARY

.DRYWELL E |||l 1i ' L v _ FUELSTORAGE

.. . . ) ~~VESSEL . .

DRYWELL WALL

:' it I e~SH l' 

;. VENT . -,WAL.... - . .. .ANNULUSJ.'

PRESSON~lMEDEST~t 5 1] HORIZONTASL

;l~~ ~~~~~~~~~~~~~~~~ E Y d N T.;11- U.

*. -CONTAINMENT

.;. . BASEMAT

- Fig. 8. BWR Mark III containment configuration. Source: Containment Performance Working Group Report, Draft Report for Comment, NUREG1037, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1985.

provided for anchorage of the hoop tendons. Hoop prestressing is provided-by 3 groups of 193 tendons, with each-group anchored at 2 buttresses spanning 120° Vertical prestressing is provided by 216 vertical

(meridional) equidistant tendons placed near the inside and outside wall

surfaces.- Dome prestressing consists of 3 groups of 63 tendons, with

each'group at 1200 with respect to the other groups. All tendons consist

-of:ninety 6.35-mm-diam steel wires and are posttensioned to -1.65 GPa.

Light reinforcement consisting primarily of Nos. 10,- 11, and 18 Grade 60
19:

ORNL-DWG 86-4560 ETD

KI

POLAR CRANE ,

STEAM

GENERATORS 

SEAL TABLEGUIDM I UvQa . - 1'Q6. _ ..

REACTOR VESSEL REACTOR CAVITY

Fig. 9. PWR large dry containment configuration. Source: Containment Performance Working Group'Report, Draft Report for Comment, NUREG1037, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1985.
20

rebars is also present. Reinforced concrete is utilized also for biological shield and' support of the reactor vessel and steam generators.

PWR hemispherical dome - three buttresses. The containment at

Trojan21 is a fully continuous reinforced concrete structure having a

cylindrical wall -. l-m-thick, a hemispherical dome 0.8 m thick, and an

-2.7-m-thick base mat-(Fig. 10). The cylindrical and dome portions of

the structure are prestressed by a posttensioning system consisting of

horizontal and vertical tendons. Three buttresses are equally spaced

around the containment. The cylinder and lower half of the dome are prestressed by 150 hoop tendons anchored 2400 apart by bypassing the intermediate buttress. Each successive hoop is progressively offset 120° from

the one beneath'. Seventy inverted U-shaped tendons continuous over the

dome are used'to provide vertical prestressing in the cylinder wall and

to provide a two-way posttensioning system for the dome. The U-shaped

tendons are divided into two tendon sets oriented 90° to each other in

the dome array, with all tendons in each set parallel to each other in

the dome and the midtendon in each set located in a diametrical plane of

the dome and cylinder. Primary prestressing for the containment consists

of one-hundred seventy 6.35-mm-diam parallel wires with anchorage provided by buttonheading. -

ORNL-DWG86-4561 ETD

FDN. MAT

Fig. 10. PWR hemispherical dome - three buttress containment.
21

2.2.2 Containment base mats

Base mats for reactor containment vessels can be either reinforced,

prestressed, or a combination of:reinforced and prestressed; however,

with very few exceptions the concrete foundation mats have been only conventionally reinforced. Design of the-base mat is complicated because it

must consider the system parameters (peak internal pressure, temperature,

environmental loads) as well as the 'influence of the soil-structure interaction. The base mat is required to support other loads, including

direct equipment loads and dead loads transmitted''through the containment

wall, primary loop compartment, and primary shield wall.

Depending on the siting conditions, the base mats may be founded on

rock, soil, or piles (e.g., piles at Point Beach, La Crosse, Fort

Calhoun, and Robinson 2). Thickness requirements of the base mats are

controlled by the concrete shear capacity, maximum allowable compressive

stress of concrete, maximum allowable!steel area, and'allowable soilbearing pressure.22 Where the containment concept 'requires a tendon

gallery for providing access to the'vertical prestressing tendon anchorages during construction and subsequent in-service inspections, the gallery can be considered as an integral'part of the base mat and encircles

it at the bottom.

Figures 3-10 present examples ofibase mat configurations that have

been used in conjunction with the various containment-concepts discussed

in Sect. 2.2.1. As noted earlier, the'base mats are'fabricated of reinforced concrete. Thicknesses 'vary according to loading and soil conditions, but in general base mat thicknesses have ranged from 2.6 m

(Oconee) to 4.1 m (Palisades)." The :base mats are circular in design and

may be >45 m in diameter. Concrete, normally fabricated from Type II

cement with compressive strengths from 20.7 to 34.5 MPal is used to fabricate the base mats. Either Grade 40 or Grade 60 steel bars, typically

ranging in size from Nos..9 to;18, are used to reinforce the base mat.

Example rebar layouts for the Indian-Point 3, Zion, and Grand Gulf containments are presented in Figs. -11-13, respectively.

2.2.3 Biological shield walls and buildings

Biological shield walls for commercial reactors are fabricated from

standard weight reinforced-concrete. Thicknesses' of the shield.walls

typically range from 1.5 to 4 m,-and the walls can either support all or

part of the reactoripressure vesseliweight. Concrete compressive -

strengths ranging from 27.6to 41.4AMPa are normally used for shield fabrication. Using-Yankee Rowe.-as-an example, the shield walls are reinforced with 139 kg of rebars per cubic'meter of concrete.23 The reinforcing steel is provided to take flexural and seismic loads that would

place portions of the wall in tension.

A shield building, or secondary containment,' is a medium-leakage reinforced concrete structure that surrounds the steel containment vessel

(see Fig. 3)..''The building is 'designed to provide (1) biological, shielding from' accident coniiditions' (2) bioiogical shielding verom.parts of th

reactor coolant system' during operation;4and (3) protection of the'con2

tainment vessel from low temperatures, adverse'atmospheric conditions,
22

ORNL-DWG 86-4562 ETD

HOOP

REBAR

SECONDARY

MERIDIONAL

REBAR SEISMIC

REBAR

PRIMARY

MERIDIONAL

REBAR '

LINER

Fig. 11. Reinforcements in Indian Point containment cylinder.

Source: S. Sharma, Y. K. Wang, and M. Reich, UMtimate Pressure Capacity

of Reinforced and Prestressed Concrete Containments; NUREG/CR-4149, BNLNUREG-57857, Brookhaven National Laboratory, Upton, New York, May 1985.

and external missiles.13 Typically the building can be a reinforced concrete cylinder with a base slab and spherical dome. Cylinder wall thickness is 0.9 m, and the dome is -0'.6 m thick. Concrete strengths used in

construction of the building range from 20.7 to 27.6 MPa. Concrete reinforcement is provided by Grade 60 rebars provided in'steel to concrete

ratios ranging from 0.003 to 0.017.

2.2.4 Auxiliary buildings

'Auxiliary buildings include functional units such as diesel generator'building, controL'room/building,' speiit-fuel' pit, fuel-handling

buildffng,'safety valve room, radioactive waste building,'and waste management'building.. Figure 14,'obtained from Ref. 24, categorizes the
23

ORNL-DWG 86-4563 ETD

HOOP

TENDON

HOOP

-REBAR

MERIDIONAL

REBAR

LINER

.. MERIDIONAL -

p- -TENDON

Fig. 12. Reinforcements and'tendons in.,Zion containment cylinder.

Source: S. Sharma, Y. K. Wang, and M. Reich, 'Ultimate Pressure Capacity

of Reinforced and Prestressed Concrete Containments, NUREG/CR-4149, :BNLNUREG-57857, Brookhaven National Laboratory,'.Upton,:New York,.May 1985.

location of the auxiliary function units.with-respect.to the reactor containment';building into three basic types,.with the auxiliary function

units either being a single continuous structure oran aggregate of'sev''

eral disjoined buildings. Infgeneral, however, these structures, are boxshaped, shear-wall buildings (see Fig. 15) constructed'of'reinforced concrete (concrete compressive.strengths ranging'from 27 to 41'MPa,'413-MPa

steel.rebar yield strength), but they may contain steel'beams (A36 structural steel) that support the floor slabs. Basic structural components
24

ORNL-DWG 86-4564 ETD

R - 67'-O"

CONTAINMENT WALL

, I R - 62'-O"

i. - - - #10 H(

. . r. SHEA

- : ;;N i #1* 

)OP REINF.

AR REINF. 2

1 TIES 3 BA

DRYWELL WALL I

. I

I i

I I

#9AND Il

R BUNDLES I

/ -- Lo 

HOOP REINF. I

RADIAL REINF.

.

R =36

CONTAINMENT

'-6' 1

. ;1j, Ii I*:.,. .4 . : - 7

I I

LINER PLATE |

| tEL.93-O"

TOP OF MAT.

i

EL. 83'-6"

BOTTOM OF MAT.

- I / # .. ., , # -.I

tLAY ER •"1

'-LAYER #2

I LAYER #3

"-LAYER 4

Fig. 13. Section through Grand Gulf containment'foundation mat.

Source: S. Sharma et al., Failure Evaluation of a Reinforced Concrete

Sark III Containment Structure Under Uniform Pressure, NUREG/CR-1967,

Brookhaven National Laboratory, Upton, New York, May 1982.

of the auxiliary function units include exterior and interior walls, base

or foundation slab, roof slab, floor slabs, and columns.

The main function of the exterior walls is to protect safety class

equipment and piping from external events such as tornadoes and tornadogenerated missiles. Typical exterior reinforced concrete wall thicknesses range from 0.45 to 1.2 m, as determined by the most severe penetrating tornado-generated'missile considered possible. Reinforcing steel

requirements are generally-based on the magnitude of pressure and energy

loads acting normal to~the walls.X

Interior wall thicknesses range from 0.3 to 1.2 m. The walls may be

constructed of reinforced concrete, concrete masonry, or heavyweight concrete. Wall thicknesses and the amount of reinforcing steel are selected

on the basis of resisting loads'resulting from internally generated missiles, equipment and pipe'supports, pressure transients, jet impingement,

thermal gradients, or rdiationi shielding-requirements.

''Baseor foundation slab thickness requirements are dependent on site

foundation conditions and-plant eismic threat. Generally constructed of

reinforced concrete,'the base or foundation slabs range in thickness from

i.8'to 8.2 m with'larger thicknesses required where the plant is located

on soft soils or piles.
25 

ORNL-DWG 86-4565 ETD

6

(TYPE I)

. . FUEL BUILDING

- 2 REACTOR BUILDING -'

3 - DIESEL GENERATOR BUILDING

4 -AUXILIARY BUILDING

.1 CONTROL A5 BUILDING

6 TRADWASTE BUILDING

E)

G)

0 1 HOLDUP TANK

2 .REACTOR MAKE-UP W

3 RFUFLING WATER T

EJ ', - | .... *,... .. 4. FUEL BUILDING

QD 10g - ,*- -- . . 5.s REACTOR BUILDING

'6 MAIN STEAM-VALVE' .E. . _ I I * * ;7t -- '., ' ' . WASTE MANAGEMENI 0 | B EMERGENCY FEEDWA

9 -WATER REUSE TANK

(E)~ ~ ~~~~~~~~1 ~ DIESEL GENERATOR I _ 1 4 r\I 11 -CONTROL BUILDING

2- 4 5 1 12-CONDENSATESTORA 0 .) 13 13 -TURBINE BUILDING

< _ 

ATER TANK

ANK

VAULT {TYPE II)

r BUILDING

*TER TANK

BUILDING

GE TANK

. .

- . .

: I ; ; 6

i I} - - ,.

I .

,: if 2

. I . , . .

(TYPE ll'

- 4- fJ - |.L| .1- STEAM-VALVE VAULT 

a/ ~~~2;REACTOR BUILDING '-'-- ' $ - ' ; ' ' t ' ' '3. ADDITIONAL EQUIPMENT BUILDING , . 4.- AUXILIARY BUILDING

5 -CONTROL BUILDING

6 -TURBINE BUILDING

Fig. 14. Generic nuclear power plant building layouts. Source:

E. Endebrock et al., Margins to Failure - Category I Structures Program:

Background and Experimental Program Plan, NUREG/CR-2347, Los Alamos

National Laboratory, New Mexico, September 1981.
26

ORNL-DWG 86-4566 ETD

ROOF BFLOOR SLAB

INTERIOR WALLS

INTERIOR WALLS ~ ~ ~ ~ EUIPEN

Fig. 15. Typical Category I structure. Source: E. Endebrock

et al., Margins to FaiZure - Category I Structures Program: Background

and Experimental Program PZan, NUREG/CR-2347, Los Alamos National Laboratory, New Mexico, September 1981.

Roof slabs are generally 0.46 m in thickness as determined by requirements to resist tornado-borne penetrating missiles. The quantity of

steel reinforcement is determined by the magnitude of pressure loads

arising from tornado winds or natural environment phenomena such as snow

or ice.

Reinforced concrete floor slabs range in thickness from 0.3 to 0.9 m.

Where a composite steel beam/concrete floor slab is-used the concrete

floor slab generally has a thickness of 0.3 m.. Slab thickness and the

amount of reinforcing steel or the size of supporting steel beams are determined by the loads supported by the floor. Occasionally, radiation

shielding requirements may dictate floor slab thickness-requirements.

Columns are used to provide intermediate supports for floor slabs

and primary support for overhead floors when walls are not available or

unusually heavy floor loads occur. The columns may be steel sections or

constructed of reinforced concrete, with the size dependent on the loading.
27

References

1. 10 CFR 50.

2. 'ACI Committee 318, "Building Code Requirements for Reinforced Concrete," ACI Standard'318-71, American Concrete Institutei,Detroit,

November 1971.

3. ACI Committee:349, "Code Requirements or'Nuclear Safety Related

Concrete Structures," ACI Standard 349-76, American Concrete Institute, Detroit, June 1976.

4. Safety-Related Concrete Structures for Nuclear Power Pants Other

than-Reactor VesseIZ and-Containments)'(for comment issue), Regulatory Guide 1.142, fice of Standards Development, U.S. Nuclear

Regulatory Commission, Washington, D.C., April 1978.

5. Staidard Format and Content of Safety Analysis Reports for Nuclear

Power Pants (LWR Edition), Regulatory Guide-1.70 (Rev. 3),O'ffice

of Standards Development, U.S. Nuclear Regulatory Commission, Washington, D.C., November 1978.,

6. R.'S. Orr and E.'G. Hess, ACI 318 or'349 for Radioactive-Waste

Facilities," Presented at Concrete and Cementitious Materials for

Radioactive Warte Management Seminar, American Concrete Institute,

New York, November 1, 1984. '

7. Joint Technical'Committee of'the ACI ad'ASME, "Code for Concrete

Reactor Vessels and Containments," ASME Boiler and Pressure Vessel

Code, ACI Standard 359-83, 'American Society of Mechanical Engineers,

New York, July 1, 1983.

8. U.S. Atomic Energy Commission, "Concrete'Containment," Sect. 3.8.1

in Regulatory Standard Review Pn,- NUREG-0800,'Washington, D.C.,

July 1981.

9. Material for Concr'ete Cntainments, Regulatory Guide 1.136,'Rev. 1,

Office of Standards Development,-U.S. Nuclear Regulatory Commission,

Washington, D.C., October 1978.

10. "Rules for Construction of Unfired Pressure Vessels;" ASME Boiler

and Pressure Vessel Code, Sect. VIII, American Society of Mechanical

Engineers, New York, '1965.

11. D. W. Halligan, "Preitressed 'Concrete Nuclear Plant Containment

Structures," J. Prestressed Concr. Inst. 21(5) (September-October

1976).

12. E. C. Silver, "Operating U.S. Power Reactors," NucL. Saf. 26(4),

520-32 (July-August 1985).
28

13. J. F. Allred et al., "Sequoyah Ice Condenser Containment Structure,"

J. Power Division 99(POl), 19-40 (May 1973).

14. Containment Performance Working Group Report (Draft Report for Comment), NUREG-1037, Office of Nuclear Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1985.

15. M. H. Shackelford et al., Characterization of NucZear Reactor Con-

.tainment Penetrations inaZ Report, NUREG/CR-3855, Argonne National

Laboratory, Argonne, Ill., February 1985.

16. C. P. Tan, A study of Design and Construction Practices of Prestressed Concrete.and Reinforced Concrete Containment VesseZs,

F-C2121, The Franklin Institute Research Laboratories, Philadelphia,

August 1969.

17. J. Kauakami et al., Reliability Assessment of Indian Point Unit 3

Containment Structure, NUREG/CR-3641, Brookhaven National Labora-

--tory, Upton, N.Y., January 1984.

18. S. Sharma et al., Failure EvaZuation of a Reinforced Concrete

Mark III Containment Structure Under Uniform Pressure, NUREG/CR1967, Brookhaven National Laboratory, Upton, N.Y., September 1982.

19.. N. W. Edwards, "Design Loads, Loading Combinations and Structural

Acceptance Criteria for BWR Containments in the United States,"

Paper J2/1, 5th International Conference on Structurat Mechanics in

Reactor Technology, Berlin, Germany, August 13-17, 1979.

20.- S. Sharma et al.,.Ultimate Pressure Capacity of Reinforced and Prestressed Concrete Containments, NREG/CR-4149, Brookhaven National

Laboratory, Upton, N.Y., May 1985.

21.. J. K. Aldorsebaes and K. Wiendner, "Containment Structure for Trojan

Nuclear Plant," J. Power Division 97(P02), 351-66 (March 1971).

22. K-Y Chu and A. A. Broussard,-"Nuclear Containment Mat Design Concepts," J. Power Division 100(POl), 85-96 (July 1974).

23. A. Wachrameeff and R. D. Chellis, "Concrete Usage in Atomic Power

Reactor Support," J. Am. Concr. Inst. 59(8), 1081-94 (August 1962).

24. E. Endebrock et al., Margins to Failure-- Category I Structures Program: Background and Eperimental Program Plan, NUREG/CR-2347, Los

Alamos National. Laboratory, Los Alamos, N.M., September 1981.
29

3. PERFORMANCE-OF CONCRETE COMPONENTS IN BOTH'

NUCLEAR AND NON-NUCLEAR APPLICATIONS

Reference 1 identifies five broad classes of;"deterioration influences" that can'impact concrete structures: (1) deterioration caused by

a change "in'environment (acid rain, sulfate-bearing groundwater, air'-pollution); (2) deteriorati6n'caused by achange in concrete properties

(long-term behavior of concrete-components); (3) short-term deterioration

(nonaging-related problems detected earlyin the service life of the

structure that would requireeither'acceptance,'repair, or removal of

service); (4) failure through human error.in'-design or"'construiction -

(problems that should be discovered'through quality assurance'programs);

and (5) deterioration'that'may result' from sudden-and/or unusual events

(serious accident or environmental type-isituations that would require an

investigation to establish structural integrity prior to a return to'service). 'Although'only' classes (l) nd_(2)- are'in a strict sense'related''

to aging 'phenomena,* the scope of this-review of concrete component performance was not restricted' to these two classes. The'sobjective was

also to obtain a broader spectrum on concrete performance so that in

cases- where problems developed, they-could be categorizedor-trended.

More specifically, these areas were'addressed: (1) concrete longevity,,'

(2) history of performance of 'concrete components innuclear-safety- >:

related applications, and (3) review of.problems 'experienced'with concrete material systems'mn both'general civil'enigineering and'niclear components. Results of this re'iew-were'then used to-provide trending-observations on'concrete'component-performance. '

3.1 Concrete 'Longevity

;Concrete has been utilized-as a construction.material for several

thousand years, probably starting with the use of gypsum mortars by 'the

Egyptians to fabricate structures such as.the Pyramid of Cheops in 3000

B.C.3 Although the longevity of concrete isattested to by the existence of the Colosseum in-Rome and the Pont du.Gard at Nimes, which is

still.capable of supporting modern road traffic after 2600 years,t

___-____x_-__I_;,___________._,__*_, ,.' I - .

*Aging, as defined in NUREG-1144,2 represents the'cumulative changes

with the passage of time'that'may'occur 'within a component 'or structure

because'of-ohe or more of 'the following facteors: (1) natural'internal :'

chemical'or physical' processes'during operation, (2) 'external 'stressors -

caused'by'storage or operating-e'ironment,'(3) service'wear including

changes' in-dimensions and/or'relativ'e positions of individual'parts or

subassemblies by operational cycling, (4) excessive testing, and (5) im-,

properlinstallation, application, or maintenance. r

'tAquaduct'and'Gard River' Bridge were built-about 2000 years ago.

The first' level of' the bridge was 'traiisformed ihto7 a''ro'ad bridge"-in the'

13th century. A:'new 'bridge wa's constructed adjacent to the' existing

water bridge in 1747.
30.

current hydraulic "portland" cement concretes have only been in existence

since 1824 when Joseph Aspdin obtained a patent. Despite the existence

of portland cement for over 160 years, relatively little documented information is available on the aging of concrete structures.* Three instances, however, have been identified in which concrete structures were

examined after an extended-period of service: (1). reinforced concrete

bridge:in Switzerland, (2) Portland Hall concrete wall in England, and.

(3) 30-year-old prestressed concrete beams.in Belgium. Also, one reference was identified in which concrete properties were determined over an

extended period-of time (50 years).

In 1889-1890 a concrete bridge was built in Wildegg, Switzerland,

using.the.Monier system (reinforced concrete arch with a37.2-m span

length).5 After 84 years-of service the bridge.was scheduled for removal,

thus permitting an.opportunity to determine-the behavior of the bridge

structure and to evaluate properties of the concrete and reinforcing

steel. Load.capacity of, the bridge was found.to be higher than anticipated with measured and calculated deflections agreeing quite well.

Fifty 5-cm-diam drill cores revealed. the concrete to have a compressive

strength of 60.8 MPa, to be well compacted, hardly damaged by frost, and

having.protected the reinforcing steel from-corrosion with most of the

reinforcement free of:.rust. Composition of the hydrated cement paste was

investigated by means of a scanning electron microscope, chemical and

thermal analyses, and X-ray diffraction analysis. Results of these

analyses. revealed that.the 84-year-old hydrated cement paste generally

appeared chemically and mineralogically.quite similar to a cement paste.

hydrated only a few-years and.thatthe cement pastewas almost completely

hydrated. The investigation.concluded that.the properties of the portland cement concrete had not been adversely affected, even after more

than 80 years of service.

A piece of portland cement- concrete was obtained for examination

from a precast concrete wall built in 1847 in front of Portland Hall,

Gravesend, Kent, England.6 Examination of one cut face of the concrete

with phenolphthalein!reveale'd that the concrete had-been carbonated to a

depth of only 5 mma , iidicatiig that the concrete interior was still

highly alkaline. Observations also sh6wed'that a coarsely ground cement

in concrete having a low water-to-cement ratio and well compacted and

cured may obtain a long-term, steady increase of'denseness, strength, and

durability due to slow continuous hydration of residual C3S and C2S.

From these results, it was concluded that it was possible to make concrete:of several hundred-years' durability..-

.:Two of the prestressed-concrete beams forming part of the Desmet

Bridge.at Ghent, Belgium,-were tested to failure under static loading

after-30 years. of-service 7 The beams were 28.8 m long,.had a flange

width-of 5.15 m,la web-0.175 m thick, and.a depth of 1.12 m. On loading,

the-safety factor of-the beams was determined to be 2.2. Concrete

*Considerable research has been conducted on concrete durability,

but current interest-is more related to deteriorating influences that can

impact the ability.of a concrete component to provide additional service

beyond-the.initial 40-year operating license of a nuclear plant.
31

strength was found to be 77% greater than its 28-d value, and the 7-mmdiam prestressing wire quality was essentially unchanged. Tests demonstrate that even after 30 years of service the beams were still in a

satisfactory condition.,

Despite the extensive amount of information available in the literature reporting results of research conducted on concrete materials

and structures, only limited data are available on the long-term (40 to

80 years) properties of portland cement concrete that are of interest

to this study. Where concrete properties are reported for conditions

that have been well-documented,-.the-results'are generally for concretes

having ages <5 years or-for specimens that were subjected to extreme,

nonrepresentative, environmental conditions such as seawater exposure.

Reference 8, however,' is'an exception because test results have been obtained from concretes prepared'under well-documented conditions for ages

up to 50 years.- In this study,:several'mortar -and concrete mixes were

prepared from a variety of aggregate and cement materials. After fabrication, the specimens weremoist cured for either 14 or 28 d and then

either stored indoors (16 to 270C),'outdoors (-32 to 350 C), or'underwater. Results obtained' from this'study showed that (1) the compressive

strength of comparable concrete cylinders stored outdoors made with high

C3S content and low surface 'area cements generally increased as the

logarithm of age to 50;years,.but concrete made with lower C3S content

and finer particle size' cements appeared to reach. peak strength at ages

between 10 and'25 years with-'some strength retrogression thereafter;

(2) concretes 'stored indoors 'exhibited little-change in compressive

strength for ages from 2 to 10 years, but thereafter showed strength increases from 30 to 70% at 50 years*, (Fig. 16); (3) modulus of rupture at

50 years was approximately.one-eighth the compressive strength for concrete stored outdoors and one-sixth the-compressive strength for concrete

stored indoors;-(4)-all-concrete stored outdoors, despite'undergoing -25

cycles of freezing;and thawing each'winter, showed remarkably good weathering qualities during the 50-year exposure period; and (5) weight and

volume changes during the 50-year storage period were small with the

largest values obtained from specimens stored indoors (1.75% weight loss,

0.059% contraction).

3.2 History of the'Performance'of Concrete Components

in Nuclear-Safety-Related Applications

As noted in Chap. 2, principal applications of concrete to nuclearsafety-related components has been in the form of containments, containment base mats, 'and biological shield walls'. Other applications'include

balance-of-plant facilities. In'the following 'sections an overview of

the performance of these components will be pre'sented. Specific items'

*Companion cylinders stored outdoors exhibited strength increases

from 10 to 40% during the 10- to 50-year period.
32

ORNL-DWG 86-4567 ETD

7,000

6.000

(400)

-j /C- 0.5 

N 5,000 E

.(300)

0.

1* 4.000

z

W

uj 3,000

> (200) 0-

C',

1,000 STORED OUTDOORS

STORED INDOORS --- b 

1923, 1:2:4 by vol., w/c 0.51 by wt.

1910,1:2:4 by vol., w/c 0.63 by wt.

1910" 1:3:6 by vol., wv/c 0.90 by wt.

0 . I1 I I " _........ I I I 

28 da 6 mo 1 yr 2 yr 5yr 10yr 25 yr 50 yr

AGE

Fig. 16. Compressive strength-age relations for Series A and B concrete stored indoors and outdoors. Source: G. W. Washa and K. F. Wendt,

"Fifty.Year Properties of Concrete," J. Am. Concr. Inst. 72(1), 26

(January 1975).
33

addressed will include: prestressed concrete containments (PCCs),* pre-,

stressed concrete reactor vessels (PCRVs), and miscellaneous reactor '-

structures. - -

3.2.1 Prestressed concrete containments

In-service surveillances ofPCCs are conducted to ensure structural

integrity and to identify'and-correctproblem areas before they become

critical. Results obtained-from these investigations are-invaluable not

only for verifying that the containments will meet their intended function, but also'from 'the standpoint of establishing performance histories.

Utilizing the component history data obtained for acontainment', aging'

trends should be relatively-easy'to-establish and should significantly

simplify the evaluation required for life extension. Information of this

type is available from surveillances of PCCs conducted in'the United

States, France, and Sweden.10

3.2.1.1 United-States.* Performance of prestressing systems has

generally been exemplary with the few document dproblemst or abnormalities being minor in nature* All: the surveillance reports concluded

that the respective containments'were'in good condition.11 ,12 Except

*for one instance in which a significant amount of water was found in several tendon ducts,§ little water-has been found during inspections.

A few instances of wire corrosion have been-reported, but these generally

'did not result-in wire breaks and were, so minor that-component replacement was not required. Thelgeneral conclusion was that corrosion had occurred prior to filling the ducts with corrosion inhibitor. Incidents of

incomplete-filling of tendon ducts with corrosion inhibitor and improper

tendon stressing have-been reported, but neither have caused-any serious

difficulties and have since been corrected. Missing buttonheads have

been discovered on some wires of buttonheaded prestressing systems;

*Only prestressed concrete containments are addressed because they'

constitute a majority of the' concrete'containments in existing' plants,

and their performance'is well-documented because'of U.S. surveillance requirements9 for the prestressing tendons. ;

-Problems documented are those that have generally been-discovered

during in-service inspections.. Problems identified during construction

or early -in the containment life -are detailed ;in Sect. 3.3.2.

*Results of a'review&1 'of the-durabilityperformance of post,

tensioning-tendons used'in conventional civil'engineiering structures-

(pavements,' bridges, etc.) have'produced'a similar'result. 'Of the over

30 million tendons'used'throughout the western world-(to 1978), the number of corrosion incidents (200 'in"completed permanent structures) represents an extremely small percentage (0.0007). All of the corrosion- -

related incidents identified were related to' -either -ill-conceived detailing, poor construction, or contaminants causing.corrosive environments.

§Despite the presence of water, corrosion was found to be very

minor,'and steps were taken'to'eliminate recurrence. This demonstrates

the effectiveness of 'corrosion inhibitors even under severe conditions.
34

however, the number of noneffective wires allowed in the design for a

tendon or group of tendons was not'exceeded. --

3.2.1.2 France. Prestressed concrete containments in France utilize grouted tendons except for four vertical tendons* in the first unit

to be built at a site. As of 1982 ten leakage and structural pressure

tests had been conducted. All leakage rates were within satisfactory

limits and the response of the containment structure was elastic and consistent with the design analysis. The few cracks that occurred during

construction were due to shrinkage and did not grow during the pressure

test.

3.2.1.3 Sweden., As of.1982, six prestressed concrete containments

were in'operation. Five of the containments utilize ungrouted tendons.

Periodic in-service inspections of .the prestressing system are carried

out mainly in accordance with.Ref. 9. For inspections reported in Ref.

10,'no serious corrosion, broken wires, or missing buttonheads have been

observed. Small amounts of water have'been found in a few grease caps

and tendon ducts, but physical tests of the grease showed that it was in

good condition, and tensile and bending tests of the wires yielded good

results. Steel properties have not been affected by time, and prestress

losses were generally less than expected.

3.2.2 Prestressed concrete reactor vessels

- As of 1982 25t PCRVs are operating, under construction, 'or planned.10

Experience from surveillance of PCRVs is available from the United Kingdom, France, and the United States.

--_3.2.2.1 United Kingdom. Checks on residual anchorage force are

made'on at least 1% of the tendons in each vessel during an inspection.

In general, no problems have been encountered with loss of-tendon load,

although--individual tendons at'Oldbury andtHinkley-have been found to be

at a lower load than expected.* Anchorage condition has been good except

some slippage was observed where tendons were removed for corrosion examination and replaced. A small number (147 out of 320,000) of missing

buttonheads were found at the Dungeness "B" vessels. Major 'corrosion instances 'occurred during'constructionwith the main causes being a combination of moisture, chloride contamination, and impressed electrical currents (probably due to improper grounding 'of dc welding machines). No

instances of corrosion were revealed that could be considered-serious

enough to warrant'tendon replacement. Extensive concrete surface exams

in which all cracks were mapped and their lengths and widths noted have

revealed all cracks'to be narrow (<0.1 to 0.2 mm), of'no structural significance, and'associated with either'drying shrinkage, construction

joints, or steel embedments. Foundation settlement has been small compared with allowable limits 'for settlement and tilt.

*Tendon ducts filled with grease.

tMarcoule G2 and G3 in France have been closed down.

*Some tendons at Hinkley were retensioned, and subsequent inspections have revealed no further significant loss of load.
35 '

3.2.2.2 France. Performance of the French PCRVs,13 of which one

has been in service over 15-years, has been satisfactory.* -Each structure is equipped with instrumentation-to monitor: forces in selected

prestressing tendons (most are grouted), horizontal outside-diameter atselected elevations; and deflections, overall tilt,:concrete temperature,

and unit deformations. Visual'inspections are-also made. Except for a

few isolated locations, temperatures have been found to be within permissible tolerances. Concrete-dimensions have tended to decrease-slightly -

because of shrinkage and-creep. Relatively few extremely fine surface

cracks are'visible. -

3.2.2.3 United States. Performance of the concrete vessel at Fort

St. Vrain has been good up to 1984 with surveillances-performed continuously since'the-initial-'proof test-in 1971. Structural response of the

vessel to pressure changes was'foundito be essentially linear, with'

strains and deflections being 'in general'agreement with those predictedby elastic analysis. A number of exceptibns'to the expected temperature

levels were found' in small areas-at discontinuities"at penetrations or -

internal-attachments to the liner caused'by shortcomings of the insula--

tion or'cooling system, but additional design assessments determined -that

these temperature levels were acceptable. However,-during'a scheduled '

1984 tendon surveillance, certain'PCRV tendons had broken, and corroded'

wires were discovered.J'+5 -Failure was caused by general corrosion and

stress corrosion cracking resulting-from the presence-of acetic and formic acids formed by microbiological attack on the anticorrosion grease.

Most of the corrosion failures were observed near the top anchor assembly

of longitudinal tendons and near the anchor assembly on'bottom crosshead

tendons. An-analysis to evaluate the integrity of the -PCRV with degraded

tendons found that the reactor vessel was capable of withstanding the operating pressures with'-the degraded tendons as determined at-that time.15

The licensee has proposed -halting degradation by filling'the tendon

sheaths with an inert nitrogen blanket and revising the surveillance program to increase the frequency of the visual inspection and lift-off

tests. The surveillance program will compare an uncorroded tendon control group'with a corroded tendon group to establish the effectiveness of

the corrosion-arresting method and the trend in tendon wire degradation.

Based on these provisions, an udated Fort St. rain Tendon Corrosion

Safety Evaluation'has recommended-that plant restart be permitted.16

3.2.3 Miscellaneous reactor structures

Probably one of the most'documented''surveys of the condition of concrete components (other than containments) in nuclear power plants that

*Extensive corrosion of longitudinal tendons in the Marcoule G2 and

G3 containments was detected during periodic surveillances conducted in

1962-63. Cause was attributed to excess humidity in conduits. Subsequent corrosion was arrested by'changing the conduit air sweeping' system

from periodic to'continuous, which maintained the relative humidity at

10%. 'Circumfere'ntial'cables'protected by several layers of bituminous

material exhibited only minor'rust. Marcoule reactors are presently decommissioned. - -
36

have been in service for several years is contained in Ref . 17, which is

related-to extension of the. service life of the Savannah River Plant

(SRP) reactors. Reactors at SRP have operated for -25. years with three

plants still operating (P, K, and C), one shutdown in 1964 R), and one

that operated to 1968.that isbeing considered for restart (L). Accumulated fast fluence in.the-reactor vessel walls was 2 x 1021 neutrons/cm2

(E > 0.1 MeV). The.tank wall temperature at full power was 100 to 1200C,

and the maximum thermal fluence was 2 x 1022 neutrons/cm2.

Reactor buildings in.the P K, andC areas were inspected for signs

of structural distress and to determine if they would support operation

for the- next-20 to:30 years. The reactor support and biological shield,

actuator towers, crane-haunches in the process rooms, and crane maintenance, areas were-inspected, and all were-found suitable for continued

support of reactor operation for-the next 20 to 30 years. Minor random

cracking was found in the P reactor building on all surfaces of the biological-shield wall. Some hairline cracks were also found running from

the wall to the edge of the.crane.haunches, and some cracks were also

noted in the-actuator tower.. In-the K building minor cracking was found

in.the wall, grouting.had failed under some of the remotely controlled

charge and.discharge-crane rail support plates, some hairline cracks were

noted in the vertical face of the crane haunches, and a vertical crack

was, found in the actuator tower. The most extensive cracking was found

in the biological shield of the C reactor building (Fig. 17). Additional

,~ . ' - >' -. . .: .¢--:,;. : -. . - ORNLPHOTO4009-86

.-

:.Fig..17. C reactor cracks in biological hield at SRP. Source:

D.. A. Ward, Extended Service Life of Savannah River PNant Reactors, DPST80-539, Savannah River Plant and Laboratory, Aiken, South Carolina,

October 1980, p. 29.
37

cracking in the C building was found beneath the anchor point of one beam

from which a piping support wassuspended in a pump room, spalled and

cracked grout was found under.one side of.,the charge and discharge crane

rail supports, and one hairline.diagonal crack was found in. the actuator

tower.

Consultants reviewed cracking in these buildings to assess the feasibility of repair. Although it was determined that the cracks were not

associated with a structural load condition, theyshould be monitored

with time to verify that they are.inactive.--Testing.of core samples was

also recommended so that the cracks.could be diagnosed more closely

(depth) and petrographic studies conducted. After' repair' of the cracks

by epoxy injection, the structure should be able to continue to meet its

functional requirements for an additional 25 years..

.3.3 Problems Experienced with Concrete Material

Systems in.Both General-Civil Engineering.-

and Nuclear Components .

Results presented-in'-the previous two-sections demonstrate that concrete fabricated from good quality materials and exposed to a normal'

atmospheric environment has indefinite'longevity and that the general

performance of con'crete'in nuclear-related applications'has been'exemplary. Problems:do occur, however, that can result in concrete distress.

To trend the type'of problems'that have been experienced with concretematerials and structures, --the literaturewas reviewedt'with respect to

bothgeneral civil engineering structures and light-water reactor (LWR)

applications. - ' ' - - - -

3.3.1 General civil engineering components

-Reference 18 presents results of 277. cases of errors.in concrete

structures obtained from a survey of consulting engineers.and.government

agencies in North' America conducted -by American Concrete Institute Committee 348. Approximately three-quarters of the 277 cases of..'error.reported were actually discovered by the structure'with 39 cases of collapse and 172 'cases of diseress,: cracking,'spalling,' leakage, settlement,

deflection,'or.rotation. '-About one-half of the errors originated in the

design and the other one-half were due to'faulty-construction with each

phase responsible for approximately.the'same number -of collapses. Design

errors were far more prevalent than construction errors in elements requiring close attention to'detail (connections, joints,-and'prestressed

members)..--Three-quarters of the:errors caused by ,faulty construction

were detected'during construction and over one-half of the errors-resulted in failure or distress. Installation of reinforcement and concreting procedures accounted for a majority of the construction errors.

Design errors,.however, weregenerally.not detected-until occupancy, with

most resulting ,iiiserviceability'problems. Design' errors resulted

largely from'improper consideration of details 'or'shrinkage and temperature effects. A'limitati6n'of the'study18'was that'the-finformation presented is strongly biased toward errors that'escaped detection until-revealed by the structure and thus does not present a true picture of the
38

error detection process of the'review-check system. Also, the survey

favored those' structures and serviceability characteristics that reveal

themselves'in a short period of time and thus does not represent the

actual incidents of concrete deterioration.'

A similar study19 reviewed -800 European failures with the focus

being on the'most efficient way' to maintain a given level of structural

safety. Results of the' survey indicated that few'structures actually

fail in use. Where-failures did occur, the type of'structures involved

included general buildings (52%), industrial buildings (22%), highway

construction (11%), hydraulic construction (7%), fallout shelters (2%),

and'unknown (6%). Primary components initiating the failure were the

structure (44%), interior works (19%), technical installations (11%),

secondary construction (9%),' construction equipment (7%), excavation pit

(5%), and unknown (5%). With respect to time of discovery of failure,

52% were discovered during construction, 45% during occupation, and 3%

during demolition. --Of the 384' cases'of structural failure,' 63% resulted

in sudden failure'(loss of equilibrium, rupture with collapse, and rupture without collapse) and. 37%-'in unsatisfactory conditions (excessive

cracking and excessive displacements). In some cases, the structure itself initiated the. failure because of unfavorable influences of the

natural environment and incorrectly introduced factors either in the

planning or construction phase.. Errors in the planning phase occurred

primarily in conceptual work or during structural analysis. Both the engineer and contractor were involved, each committing errors because of

insufficient knowledge or ignorance. Only very few errors were unavoidable, and in a majority of cases additional checking; would have helped

considerably. From these results it was concluded that a primary deficiency in structural safety was attributed to insufficient data checking.

3.3.2 LWR concrete components 

Results presented"in Sect. 3.2 indicate that in general the inservice performance of concrete materials and components: in nuclearsafety-related applications has been very good. This to a large degree

can be attributed to the effectiveness of the quality control/quality assurance programs20 in detecting potential' problems. (and'the subsequent

remedial measures)'prior toplant operation.*, To obtain information on

the type of 'problems that have been experienced (detected) with LWR components, computer searches. ofi Licensee Event Reports (LERs), the Nuclear

Plant Reliability Data System-(NPRDS), the Nuclear Power Experience (NPE)

data base, and Construction Deficiency Reports (CDRs) have been conducted. Information- has also'been obtained from the DOE/RECON computerized 'information' retrieval system and appropriate journals (Prestressed.

Concrete Institute, American Society of Civil Engineers, Engineering

*Additional information on the'effectiveness of structural concrete

quality 'assurance practices in nuclear power plant'construction 'is presented in' Ref. 20, which reviewed nine nuclear 'and three fossil fuel

plant construction'projects.'
39

News-Record, etc). Obviously, because of the sheer magnitude of documentation available for even one plant, all problem areas were not'determined. However, the anomalies identified are characteristic of those,

problems that occur and thus rovide trending information.

Figure 18 summarizes results of the survey according to'problem type

(concrete cracking, tendon failure, etc.),'and an annotated problem listing is presented in Appendix B. The majority of problems weie-related

to concrete cracking, concrete voids,' or honeycombing, and concrete compressive 'strength values that were 'low relative to design values at' 'a

specific concrete age. In almost all cases, the concrete-cracks were

considered to be structurally insignificant or easily repaired using

techniques such as epoxy' injection. Voids and honeycombed'areas were

restored by'removing faulty materials and making repairs using acceptedprocedures such as grout injection', drypacking, or shotcreting. In a few

instances low-strength concrete materials had to be removed and replaced,

but in the majority of situations'either the in-situ strength'was''determined to be in excess of design requirements or subsequent tests conducted at later concrete ages achieved acceptable strength levels.

Although the vast majority of the problems detected did not present'

a threat to public safety or jeopardize the structural integrity of the

particular component, five incidences were identified that if not discovered and repaired could potentially have had serious consequences.

These incidences were all related to the'concrete containment and involved two dome delaminations, voids under tendon bearing plates, anchor

head failures, and a breakdown in quality control and construction management. Note that these incidences were attributed either to design,

construction, or human errors, but not-to aging.

After 110 of 165 tendons 'in the containment dome of Turkey Point 3

had been tensioned, it was noted that sheathing filler was leaking from a

ORNL-DWG 86-4568 ETD

. . . . : NUMBER OF IDENTIFIED OCCURRENCES|

CONCRETE (77) 0 5 10 15 20 25

VOIDSIHONEYCOMB

CF;ACKING/SPALLING

DEFECTIVE MATERIAL/LOW fi

IMPROPER PLACEMENT/REPAIR

ANCHORAGE

REINFORCING STEEL

OVERTEMPERATURE . '. .

DOME DELAMINATION

POST.TENSIONING SYSTEM (17)

FAILED/CORRODED TENDONS

ANCHORAGE CRACKING

LIFTOFF LOAD

Fig. 18. Distribution of LWR concrete component problem areas.
40

crack'in the dome- surface.21 A small amount of concrete was chipped away

adjacent to the crack to reveal a crack plane parallel to the surface

(delamination) with evidences of sheathing filler'flow on the delaminated

surfaces. Five days later, a smallibulge was nted in the dome surface,.

which when broken through revealed a delamination at a depth of about

12.7 mm.. Exploratory chippingrevealed that the delamination became

thicker as the dome center was approached, reaching'-102 mm when chipping

was terminated at a 4.6-m radius. Soundings were taken with a Swiss hammer and steel sledge hammer-to indicate the area affected. Sixty-five

102-mm-diam cores were' drilled to estimate the depth and extent of delaminations. The core' samples revealed'that the depth'and extent of delaminations was considerable and symmetrical, the delaminations appeared

to have originated at ameridional 'construction joint, and many of the

cores had sheathing filler in them as well as showing signs of multiple

delam'inations. Dome tendons, of which all but two had been tensioned at

this time, were detensioned. The delaminated concrete was removed by

chipping guns and jack hammers. As shown in Fig. 19, the'delaminations

covered >50% of the dome and reached depths to '-0.4 m. Exposed concrete

ORNL-DWG 83-8792A

CONSTRUCTION JOINT

0 -0 ° 90° 2700 + ' 0-15 900

1800 CONTOURS OF DEEPEST

DELAMINATION

I 1 II

0 Sm

SCALE

Fig. 19. Extent of dome delamination for Turkey Point Unit 3.
41

surfaces were cleaned using a high-pressure air-water .blast technique.,--A

system of radial rock anchors was installed to assist in providing radial

forces on the replaced concrete. The concrete was-then replaced using

extreme care to ensure good:bond with the existing concrete. Tendons

were retensioned.using a new sequence to reduce.unbalanced loadings during the prestressing operation. An ensuing structural integrity test revealed no recurrence of delaminations. The cause of-the delamination was

determined to-be attributed to insufficient contact area:in the southern

portion of the meridional.construction joint and around the ventilation

blockouts, together with unbalanced posttensioning loads.,

Delamination of-the containment dome also-occurred at Crystal River

Unit 3.22 Discovery of the-delamination occurred. 2 years after completion-of concrete placement and-l year-after tendon tensioning when electricians could not-secure some'drilled-in anchors to-the top surface of

the dome. Further investigation revealed-an area of dome concrete that

sounded hollow when struck with a hammer. Exploratory.holes were then

cored and concrete samples.removed. Results of this investigation revealed -that the-dome had delaminated.over-an area having a diameter of.

-32 m (Fig. 20) and had a maximum thickness of delamination of 0.38 m'

ORNL-DWG 83-8793A

i, , - . 0° 360--

.- ~ .~ -,DLA ~ I AI

/ / INITIALLY - - o

DISCOVERED DIM ' . N IN DETI

I d & 'AR EA . ^\3 90° 1. 71 . 1. I I 2700

SAW CUT AR EA ..t 

.. ~ ~ ~~# ByDELAMAI.NATIONIO7 : / .

- - , - PLAN VIEW . .at

INITIAL INVESTIGATION DETAILS ' r-Ae''

W .\DIMENSIONS INcm' DETAIL A,.

Fig. 20. Extent of dome delamination for Crystal River Unit 3.
42

near the apex with-a gap of -51 mm between layers. Analysis of the

structure indicated that it was acting as a 0.62-m prestressed concrete

dome having a 0.31-m unstressed concrete cap and that although it was

safe for any normally'anticipated loads, it would require repair to withstand accident conditions. While the delaminated cap was still in place,

-1850 radial holes 25.4 mm in diameter were drilled into the dome to provide-a means for further inspection, to serve as grouting and venting

holes, and to provide access for placing radial reinforcement. The delaminated'cap was then removed, and cracks were repaired by'pressure injecting a low-viscosity epoxy. Nonprestressed meridional and hoop reinforcement was provided to enhance the membrane and tensile 'capacity of the

structure and to'control cracking. Concrete materials for the new cap

were the same s'those in-the original concrete. After concrete placement and curing, 8'tendons that had been detensioned to obtain strain

and deformation data were retensioned, and a structural integrity test

was successfully conducted. 'Based on analytical and experimental evaluations, it was concluded'that radial tension stresses combined with biaxial

compression stresses initiated'laminar cracking in the concrete that had

lower than normal tensile strength and limited crack-arresting capability.

At Calvert Cliffs nuclear plant during posttensioning, 11 top bearing plates of the 204 vertical tendons of Unit 1 containment and 1 bearing plate of Unit 2 containment depressed into the concrete.23 Ten of

the Unit 1 plates had depressed from -0.8 to 4.8 mm with the depressions

generally on the inside plate edge. However, when the last tendon was

stressed the upper bearing plate deformed and sank -25.4 mm along its inside edge. Eight months later this upper bearing plate was removed to

reveal that the plate was'supported on the outside edge by concrete occupying only 20% of'-the total area and on the inside'edge by the upper

reinforcing bar, which had deflected -12.7 mm. The revealed void was

deepest (305 mm) next to the trumpet and extended outside the bearing

plate area. Tension was then released on one of the tendons whose plate

had depressed 4.8 mm and the plate removed to reveal'concrete only in

contact over about one-third the plate area. The void was deepest (178

to 203 mm) adjacent to the trumpet. At this point, it was decided that a

comprehensive investigation of all tendon upper-bearing plates was required. Using sound-transmission, probing, and air.pressure techniques,

-190 bearing plates on each unit were identified as possibly having

voids. Affected tendons were detensioned to about 3.4 MPa and voids repaired by either pressure grouting or flow grouting. After repairs a

number of the plates were examined by drilling and probing or using air

pressure to determine if grout had been placed in the area under the tendon shims. During'subsequent structural integrity testing of Unit 1,

dial gage micrometers were used to verify that plates were rigid up to

1.15 times the containment design pressure.

Anchor head failures have occurred at Bellefonte,24 Byron,25 and

Farley Units 1 and 2 nuclear plants.*26 ,27 The failures at Bellefonte

occurred in eight of the top anchor heads of 170-wire rock anchor tendons

just prior to a two-stage grouting process used to anchor the tendons to

*Tendons and anchor heads for all three plants were supplied by the

same vendor.
...

43

the rock., In one of the anchorhead.,failures 23 of 170 wires in the tendon also failed. Environmental, metallographical, and fractographical

studies indicated that the failures were the result of stress corrosion

cracking of highly stressed AISI 4140.anchor heads in an aqueous environment of varying pH levels. Also between first- and-second-stage grout- ing, the top anchor heads were-covered.by grease cans filled with.lime

water having a pH of ll to 13. Anchor heads have been replaced with

cleaner steel and other improvements have been made. At Byron four

anchor,heads of 179-wire. tendons failed betweenl 1and 64 d after post- tensioning.the-Unit 1 containment.. A thorough study of the chemistry,

metallurgy, and fracturephenomena indicated that the failure wasp caused

by tempered martensite embrittlement (vanadium grain refinement process

used with temperatures not high enough),and occurred in a decreasing

stress field. Anchor head failures at Farley Units 1 and 2 are of recent

vintage and unique from the standpoint that the failures occurred about

8 years after posttensioning rather.-than during construction when failures

are most likely. Using magnetic particle testing, cracks in 6 anchors in

Farley 1 and 18 anchors in Farley 2 were discovered in addition to the 3

anchors that had already failed in Farley 2.* Laboratory tests have con- cluded that the ca6se of ''tendon ach-r ead failures 'was stress corrosion

crackiig caused by a combination of high-trength low-alloy steel under

high stress in the presenceof moisture andimpurities. (Inspections re- vealed that althogh only'slight amounts of 'water we found in three

hoop andone dome anchor, 47 of 103'vertical tendons were found with

water ranging from trace amounts to 5.7 L.) -All cracked and failed

anchors have been zeplaced and grease has been applied using an improved

procedure to preventth'e water-caused problem fromrecurring. In addition all tendons from 'the same lot as the failed tendon have'-been 'inspected, and 20 of the vertical tendon's hive been replaced with a superior heat material. , Safety-related concrete work at Marble Hill Unit's 1 and 2 was halted

by theNRC because of quality 5ntrol'and construction management inadequacies.28 Reasons for the halt in construction were (1) an excessive

amount of honeycomb 'and air oids with -4000 patches existing ranging in

sizeup to several squaremeters inarea, (2) imperfections in many-in-"

stances had been improperly repaired and/orunacceptable materials utilized, (3) quality control records traceableto repairs were either nonexistent orotherwise inadequate, (4)'personnel responsible for repairs

were inadequately trained and supervised, 'and (5) the licensee was

neither in control n6r'sufficiently awareofthe-above circumstances.

All patches were required by NRC to be removed and repaired, and -indepen- dent consultants;were retained to'prbvide anasessment of-the tye and

extent of deficiencies inenc' rete onstruction toprovide an assessment

of any;needed.repairsor-remedial actions, and :to provideconclusions regarding the capability,,of the,affectedstructure toperform its intendedfunction. Basedo ;othe-independent-consultant!s investigation, it was

concluded that the Marble Hill -concrete structures were constructed of.

*Design margin is not an issue because the containment has a margin

permitting-failure of several -tendons. What is important -is establishing

the cause and preventing subsequent failures. -
I

44

high-quality (strength) concrete materials, but with concrete placement

that' failed to meet requirem'ents'near the surface due to difficulty in

properly vibrating the concrete between the form and reinforcement layer.

Nondestructive testing results and coring indicated that the internal

concrete consisted of homogeneous concrete and was acceptable-.- Methods

used to detect defective concrete and concrete patch areas were consistent with good construction practice'. Investigation of-a selected number

of visible surface defects verified-that procedures crreiitly'being used

to prepare surfaces and repair areas are consistent with good construction-practice. Furthermore, it was concluded that if the repair procedures reviewed were followed and high standards of workmanship maintained, the structural integrity and biological shielding requirements ofthe concrete structures should be met.

- 3.4 Trending Observations on the Performance

of Concrete Components -

'Although the data base evaluated in the'review was somewhat limited,

results obtained are considered to be sufficiently representative that

some general observations can be made on concrete aging and component

performance. When-concrete is fabricated-with close attention to'the

factors shown in Fig. 21 (Ref'. 29) related 'to the production of good concrete, the concrete will have infinite durability unless subjected'to extreme external influences* -(overload, elevated temperatures, industrial

liquids and gases, etc.).' Under normal environmental conditions aging

of concrete does not have a'detrimental effect on its strength for concrete ages to at least 50 years.t Review of the performance of concrete

components in general'civil engineering structures indicates that few

structures actually fail in use and that the errors* that do occur are

predominantly detected during' construction'. The source of these errors

is generally the result' of either construction or design detail errors.

The overall performance of concrete components in nuclear applications

has' been very 'good. With the exception of the anchor'head failures at

Farley 2, errors detected-during the construction phase or early in the

structure's life wre of no structural'significance or "easily" repaired

and were nonaging related. ,.The 'rigorous in-service inspection programs

required of'nuclear components are achieving their desired objective of

uncovering and correcting potential problem areas and provide a valuable

*Chapter-4 presents, a discussion of environmental stressors and

aging factors to which concrete components may be subjected.

- tA limit on age for'which well-documented data has been identified.

The'number of concrete structures'in existence having ages of 40 to 70

years', ith a few iri'service'"for'thousands-of years, indicates that this

value is conservative. 'Also, many structures continue to meet their

functional and performance requirements even when conditions are far from

ideal.

-:*Errors could be significantly reduced by additional quality assurance/quality control procedures.
45

ORNL-DWG 86-4569 ETD

- , | ~GOOD CONCRETE |

SELECTION OF CONTROL OF DESIRABLE

MATERIALS PRODUCTION PROPERTIES COST

KIND BATCHING WORKABILITY MATERIALS

(QUALITY) MIXING UNiFORMITY LABOR

AMOUNT TRANSPORTING STRENGTH EQUIPMENT

(PROPORTIONS) PLACING DURABILITY,

WATERTIGHTNESS UNIFORMITY FINISHING VOUE

LFCURING CONSTANCY.

DESIRED WEI GHT

ETC.

* . CONTROL TESTS AND INSP' TIO

'Fig. 21. Factors inproduction of good concrete Source: ^G. E.

Troxell et al., Composition and Properties bf'Concrete, 2nd ed, McGrawHill Book Co., New York, 1968.

source of data for trending component performance. In the one example of

a nuclear plant that was identified where component life extension was

being c6nsidered, the main distress of concrete components identified was

cracking. Once the cracks were repaired with a procedure such as epoxy

injection, it was felt that the structures should be able to meet their

functional requirements for at least an additional 25 years.-

References

1. C. A. Negin et al., Extended Life peration of Light Water Reactors:

:Economic and Technological .Review, EPRI NP-2418, vols. 1 and 2,

Electric Power-Research'Institute, Palo';Alto, Calif., June 1982.

2. -B.-M.-Morris and J. P. Vora,-NucZear PZant Aging Research YNPAR)

Program Pan, NUREG-1144i Division of Engineering'Technology,'Office

of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

*Washington;'D.C.,'July 1985. ' - -

3. S. Mindess and J. F. Young, Concrete, Prentice-Hall, Englewood

Cliffs, N.J., 1981. s ad,. C u .. .;.. ,, . ;, .. * 

4. -P. E.dlalstead,"Causes of Concrete Decay," The ConsultingEng.

-(Supplement) (April-May 1971). . '-> *-' -

5. J. Gebauer and A. B. Harnik, "Microstructure and Composition of Hydrated Cement Paste of an 84 Year Old Concrete Bridge Construction,"

Cem. Concr. Res. 5(2), 163-9 (March 1975).
46

6. G. M. dorn and N. Thaulow, "Examination of 136 Year Old Portland

Cement Concrete," Cem. Concr. RAes. 13(5), 739-43 (September 1983).

7. F. G. Riessauw and L. Taerwe, "Tests on Two 30-Year-Old Prestressed

Concrete Beams," J. Prestressed Concr. Inst. 25(6), 70-72 (NovemberDecember 1980).

8. G. W. Washa and K. F. Wendt, "Fifty Year Properties of Concrete," J.

Am. Concr. Inst. 72(1), 20-28 (January 1975).

9. Inservice SurveiZZance of Ugrouted Tendons in Prestressed Concrete

Containment Structures, Regulatory Guide 1.35 (Rev. 2), Office of

Standards Development, U.S'. Nuclear Regulatory Commission, Washington, D.C., January 1976.

10. Technical Report - An International Survey of In-Service Inspection

Experience with Prestressed Concrete Pressure VesseZs and Containments for Nuclear Reactors, FIP/3/5, FP Commission on Concrete

Pressure and Storage Vessels, Federation Internationale de la Precontrainte, Wexham Springs, Slough, England, 1982.

11. M. Schupack, "A Survey of the Durability Performance of PostTensioning Tendons," J. Am. Concr. Inst. 75(10), 501-10 (October

1978).

12. J. R. Dougan, Evaluation of Inservice Inspections of Greased Prestressing Tendons, ORNL/TH-8278, Union Carbide Corp. Nuclear Div.,

Oak Ridge Natl. Lab., September 1982.

. G,. .oilt -In

13.- G. Douillet, 'Inspection of Prestressed Concrete Pressure Vessels of

Natural Uranium Gas-Coled Reactors Owned by Electricite de France,"

pp. 400-17 in Specialists' Meeting on Design, Criteria and Experience with Prestressed Concrete Reactor Pressure Vessels for GasCooled Reactors, International Atomic Energy Agency, Lausanne,

Switzerland, Dec. 4-7, 1984.

14. D. J. Naus,i'Review and Current Status of Fort St. Vrain with Par-

,ticular Reference tolItslPCRV," Specialists' Meeting on Design, Criteria and Experience with Prestressed Concrete Reactor Pressure Vessels for Gas-Cooled Reactors, International Atomic Energy Agency,

Lausanne, Switzerland, Dec. 4-7, 1984.

15. J. Eggebroten, Corrosion of CRV Tendon Wires, Licensee Event Report

267-005 (Rev. 3), Public Service Company of Colorado, Denver, July

5, 1985..

16. "Fort St. Vrain Tendon Corrosion-Safety Evaluation (Updated)," memorandum from H. L.-Thomson, Jr., Division of Licensing, Office of Nuclear Regulatory Regulation, to R. P. Denise, Division of Reactor

Safety and Projects (Region IV), U.S. Nuclear Regulatory Commission,

July 2, 1985.
47

17. D. A. Ward, Extended Service Life of Savannah River Plant Reactors,

DPST-80-539, Savannah River Plant and Laboratory, Aiken, S.C.,

October 1980.

18. J. Fraczek, "ACI Survey of Concrete Structure Errors," J. Am. Concr.

Inst. -1(12), 14-20 (December 1979).

19. R. Hauser, "Lessons Learned .from European Failures," J. Am. Concr.

Inst. 1(12), 21-25 (December 1979). -

20. J. H. Willenbrock et al.,- Final Summary Report: A Comparative 

Analysis of Structural Concrete-Quality Assurance Practices offNine

Nuclear and Three Fossil uel 'PowerPlant Construction Projects, Re-

- port No.-COO/4120-3, Department of Civil Engineering,'The Pennsylvania State University,' University Park, Pa.- December 1978.-

21. Containment Dome Report Turkey Point Unit 3, NRC-PDR Docket'No.

50-250, Florida Power and Light Co., Miami, Dec. 23, 1970.

22. F. L. Moreadith and R. E. Pages, "Delaminated Prestressed Concrete

Dome: Investigation and Repair," J. Struct. Eng. 109(105), 1235-49

(May 1983). -

23. Study Report on Vertical Tendon Bearing Plates, Appendix'50 of Cal-'

vert Cliffs Nuclear Power Plant Preliminary Safety Analysis-Report,

Baltimore Gas and Electric Co., NRC-PDR'Docket Nos. 50-317 and

50-319, July 1974.

24. W.-E. Berry et al-., Examination of the Cracked Rock Anchor 'in the

TVA Bellefonte Nuclear Power Pant,-Battelle Columbus Laboratories,

Columbus, Ohio, 1976.

25. S. E. Presswalla, Report on the Failure Investigation of Post-

.tensioning Anchorheads Used in the Byron Nuclear Containment Structure, NRC-PDR Docket Nos. 50-4547 and 50-455, Inryco, Melrose Park,

Ill., 1980.

26. C. Hudgins, "The Cause of Tendon Anchor Failures at Alabama Power

Co.'s Farley-2," Nucleonics Week 26(9), 4 (Feb. 28, 1985).

27. C. Hudgins, "Farley Tendon Problem Blamed on Water as NRC Mulls

Generic Impact," NucZeonics eek 26(10), 1 (Mar. 7, 1985).

28. A. L. Parme and R. C. Hamm, Review of the EvaZuation-of Concrete-at

-Marble Hill Nuclear Generation Station nits -1 and 2, Report No.

IE-124, NRC-PDR Docket Nos. 50-546 and 50-547,'Public Service Co.-of

* Indiana, New Washington, iId., April 28,' 1981. ; '

29. G. E. roxell et al., Composition and Properties of Concrete, 2d

ed., McGraw-Hill Book Co., New York, 1968.
48

4. POTENTIAL ENVIRONMENTAL STRESSORS* AND AGING FACTORS

TO WHICH LWR SAFETY-RELATED CONCRETE

COMPONENTS MAY BE SUBJECTED

Reactors are generally designed for a plant life of about 40 years,

which, with an anticipated availability factor of 80 to 90%, yields 32

to 36 full-power year's. Over this period of time, changes in concrete's

material or reinforcing steel properties in. all likelihood will occur as

a result of environmental effects such as elevated temperature or irradiation. As noted in the Chap. 3, the changes in properties do not have to

be detrimental to the point that the structure has deteriorated and is

unable to meet its functional and performance requirements. This is also

pointed out by Mather,1 who notes that when the specifications covering

concrete's production are correct and are followed, concrete will not deteriorate. Concrete in many structures can suffer undesirable degrees of

change with time because of improper specifications or a violation of

specifications. Mechanisms (factors) that, under unfavorable conditions,

can produce premature concrete deterioration include (1) freezing and

thawing, (2) aggressive, chemical exposure, (3) abrasion, (4) corrosion of

steel and other embedded material, (5) chemical reactions of aggregates,

and (6) other factors (unsound cement and shrinkage cracking). Table 3

(Ref. 1) presents concrete characteristics, environmental characteristics, and the manifestation of deterioration for each of these factors.

For concrete components utilized in nuclear-safety-related structures, an

additional factor can be added, extreme environmental exposure (e.g.,

elevated temperature and irradiation).

In nuclear-safety-related concrete components, the relevant degradation factors that can influence component performance vary by application.

Potential degradation factors for reinforced concrete containments (RCCs)

are related to those that cause deterioration of the concrete or reinforcing steel. For prestressed concrete containments (PCCs), the factors

would be the same as for RCCs except that factors that would cause deterioration of the prestressing system would have to be added. Factors

affecting containment base mats would also be the same as.-those for RCCs,

plus those contributing to foundation settlement and aggressive chemical

attack by the groundwater. Biological shield walls would be susceptible

to factors that would produce a loss of concrete strength or shielding

efficiency. Table'4 presents a' summary of the predominant environmental

stressors to which-safety-related components in a light-water reactor

(LWR) plant could be subjected that may cause an effect leading to deterioration (nonaccident'conditions). In the following sections potential

deterioration of these components is discussed in terms of factors that

*An extreme load condition-' is not considered because it is not an

aging-related occurrence. If an overload condition occurred, it would

be a readily discernible event and require-detailed structural inspection

and evaluation.
49

-Table 3. Interacting factors fr mechanisms producing - . I -,. premature concrete deterioration 

Factor that may -- indcer-ehat may Characteristic of Characteristic of Manifestation of

premature thelconcrete. the environment deterioration

deterioration - - I 

Freezing and

thawing

Aggressive chi -n v _ 

Lack of entrained air Moisture a

in the ciement paste and thawi

or excessively porous -

aggregate,'or both,

in saturated concrete

emit '

and freezing' 'Internal'expansion'

ng -; - and cracking

Cal LLac

Sulfate

Leachir

Abrasion

attack Excessive amounts o

hydrated calcium

aluminates in the

cement paste

Ig ! Excessive porosity

T 

Lack of'resistance

abrasion

f Moisture containing

dissolved sulfates '

in excessive concentration

Moisture of low pH 

and low dissolved

lime content

to Abrasive, often In

or under-water

Internal expansion

and cracking

Dissolution and

removal of soluble' constituents

Removal of material.- -

Corrosion

embedded

Alkali-si:

reaction

Other

Unsouni

cemen

Plasti

shrin

crack

of Corrodible metal-and Moisture (or mois- Internal expansion

metal- (usually) corrosion- t- ure and corrosion- - and cracking

- inducing agents in --inducing agents)

*the concrete

lica --Excessive amounts of Moisture (or mois- * Internal expansion

soluble silica in the ture and alkalies) and cracking

aggregate and (usu- ' -

' ally) alkalies in

the cement '' '

Id

t

c

kage

ing

Excessive amounts of Moisture

unhydrated CaO of

MgO in the cement

Lack of maintained High evaporation

moisture content dur- rate for moisture

ing specified curing.-

period

Internal expansion

and cracking

Cracking at very

early ages

Sourcei B.-Mather, "Concrete eed Not Diteribrate," J.-Am.' Conc Inst. 1(9),

33 (September 1979). ' . '

can affect the durability of the materials'used to fabricate the compo-'

ne'nts,-that s, concrete, reinforcing stee prestressing tee 

anchorage embediments.* ' ' se '' t a''

*Although anchorage embedments are not a constituent'of concrete

components per se, they'must function with the concrete within.which they

are embedded.
50

Table 4. Predominant environmental stressors to which

safety-related components in a LWR may be subjected

StruCtural Important Predominant

material environmental subsystem components - a parameters stressorsb

Prestressed concrete Concrete fc, Ec, v, CR D, L

containment Rebars fy, E8 , q,, e C, L

Prestressing fy, E5, au, R C, L

Reinforced concrete Concrete fA, EC, V D, L

containment Rebars fy, Es, a, e C, L

Containment base mat Concrete f, EC, V D, L, S

- Rebars fy, Es, au, e C, L

Biological shield wall Concrete fA, EC, V T, I, L

or building Rebars fy, Es, a, e T, I, L

Prestressing fy, Es, a T, I, L

Auxiliary buildings Concrete fc, Ec, V D, L

Rebars fyp Es, a,, e, C, L

af A

E

V

CR

au

e

R

fy .y

= concrete compressive strength

- modulus of elasticity

- Poisson's ratio '

- concrete creep

- ultimate strength

- elongation or ductility

- prestressing relaxation

= steel yield strength

bT . temperature

D = durability

I - irradiation

C - corrosion

L - external, internal,

or'dead loads

S = subgrade settlement

4.1 Concrete Degradation

Concrete is a general term for a class of ceramic materials that

vary widely in their properties and applications. The American Concrete

Institute (ACI) defines concrete as "a composite material that consists

essentially of a binding medium within which are embedded particles or

fragments of aggregate; in portland cement concrete the binder is a mixture of portland cementand water."2 By varying the constituents and

their relative proportions in the mixture,. concretes of widely differing

properties can be obtained, fr example, strengths from 0.7 to 100 MPa

and unit weights from 800 to 4000 kg/m 3. Concrete materials utilized in

LWR applications generally have compressive strengths ranging from 20.7

to 41.4.MPa and unit weights from 2240 to,2400 kg/m 3. Potential causes

of deterioration of concrete would be cracking, aggressive environments,

embedment corrosion, or extreme environmental exposure.
51

4.1.1 Concrete cracking

Cracking.occurs in virtually all concrete, structures and,-because of:

concrete's inherently low tensile strength and lack of ductility, can

never be totally eliminated. Cracks are significant from the standpoint

that they can indicate major ,structural problems (active cracks); provide

an important avenue for the ingress of hostile environments,(active or:--.

dormant cracks); and inhibit a component from meeting its performance requirements,.such as providing biological shielding (active or dormant

crack).. As.noted in Table 5 (Ref. 3), cracking results from each material component and can occur-while the concrete is in either a plastic or

hardened state.

Table 5. Causes of concrete cracking

Component ' Type Cause -of Environmental ' Variables. -.p.e.Type distress I. factor(s)', to control

Cement Unsoundness Volume . Moisture. Free lime and magnesia

expansion. .

TemperatureTemperaturemperature Heat of hydration,

cracking stress ' rate 'of cooling'

Aggregate Alkali-silica Volume Supply of. Alkali in cement,

reaction expansion' moisture composition of

'-' ' ; '- ' '.'' 'aggregate

Frost attack Hydraulic Freezing and Absorption 'of 'aggrepressure . . thawing .gate, air content

of concrete, maximum

size of aggregate

Cement - Plastic . Moisture Wind and Temperature of concrete,

paste shrinkage loss . temperature protection of surfaces

Drying Moisture . Relative Mix design, rate of

shrinkage loss humidity drying

Sulfate attack Volume Sulfate ions Mix design, cement

:expansion ' ' type, admixtures

-Thermal . . Volume, Temperature Temperature rise,

expansion . .expansion' ichange rate of change

Reinforce- Electro- * i' Volume -Oxygen Adequate concrete

ment . chemical - expansion -moisture cover

corrosion ' -

Source: S. Mindess and J. F. Young, Concrete, Prentice-Hall, Inc., Englewood

Cliffs, New Jersey, 1981, p. 572. .. -.

4.1.1.1 Cracking of concrete during initial setting (plastic concrete cracking). Cracking of concrete -during.initial setting can result

from a number of causes: (1) settlement due to unstable subgrade;

(2) poor form construction; (3) lack of, insufficient, or improper rebar

placement; (4) rebar crrosion; (5) high-slump concrete; (6) improper
52

consolidation; (7) lack of curing; (8) volume change due to solids settlement; (9) insufficient expansion or control joints; and (10) early stripping of forms.4 Most cracking of plastic concrete, however, occurs in

three-primary forms: plastic shrinkage cracking, settlement cracking,

and crazing.

Plastic shrinkage cracking occurs most frequently on the exposed

surfaces of freshly placed floors'and slabs subjected to a rapid loss of

surface moisture caused by low humidity,- in&,' or high temperature. The

cracks form'as a result of differential volume change in which concrete

near the surface tries to shrink but 'is restrained by the concrete below.

Shrinkage cracks can range from a few millimeters to several meters in

length, with spacings-from a few millimeters to >3 m; and although generally shallow, they can penetrate the full depth of an elevated slab.

Cracks of this type are not a direct threat to the structural integrity

of a member but indirectly can have an effect if they are of sufficient

width to permit entry of a hostile environment.

Settlement cracks result from constraint provided by obstacles such

as reinforcing bars or other embedments, form work, or a prior concrete

placement. These cracks do not occur where the foundation was properly

prepared, forms were properly designed, the ix was properly proportioned

to have the lowest possible slump consistent with placement, and adequate

concrete compaction was'provided.5 Settlement cracks have the same effect on a structure as plastic shrinkage cracks.

Crazing sometimes occurs in a hexagonal pattern on the concrete surface at an early age because of improper curing (excessive water loss)

and finishing procedures (excessive'flotation or troweling). Crazing is

surface related and thus not a threat to the structure.

4.1.1.2 Cracking of hardened concrete.* Cracking of hardened concrete results from shrinkage with restraint, thermal effects, and chemical reactions.t

Drying of hardened concrete. Concrete contracts (shrinks) as it

loses water; if the concrete is constrained cracking can occur when the

tensile strength of 'the concrete 'is exceeded. Factors that affect the

volume change of mortars or concretes caused by variations in moisture

conditions include: (l) cement and water contents, (2) composition and

fineness of cement, (3) type and gradation of-aggregate, (4) admixtures,

(5) age, (6) test duration, (7) moisture and temperature conditions,

(8) specimen size and shape, (9) form absorption, and (10) amount and

distribution of reinforcement.6 Cracking due to volume change not only

may impair the ability of a structure to carry its designed loads but

also may affect its durability and damage its appearance.

Carbon dioxide, present in-the atmosphere, under some conditions may

react with the Ca(OH)2 or other lime-bearing compounds in hardened concrete to produce a reduction in volume and an increase in weight. Crack--

ing resulting from carbonation is generally confined to a thin layer near

*Although not considered in this section, primary concrete containments' can develop cracks during;structural acceptance testing when the

containment's internal pressure is increased to 1.15 times the design

pressure.

tEffects of embedment corrosion including concrete cracking are

covered in Sect. 4.1.3.
53

the surface. Another effect of' carbonationis that in the areas where

calcium carbonatezforms, the pHof the'concrete is lowered 'sufficiently

(8.5 to 9.0) to destroy the passivating'effect'of the concrete on rebars,

which potentially enables the rebars to corrode.'i'Fortunately the car- -

bonation process is slow, and its 'rate'is inversely proportional'to concrete quality, so it should have minimal effect on mass concrete structures. . . .

Thermal effects resulting from;cement hydration. The'setting and

hardening process of concrete is a chemical reaction that liberates heat

on the order of 60 to 120 cal/g of cement.6 If the heat cannot be'dissipated to its surroundings fast'enough,' a temperature rise of 40C or

greater "can occur,3 and the mass'will also expand. During cooling', the

outer concrete surface' cools first and shrinks,''which can result in the

formation of cracks. Because of the 'constraint provided by' the'inner concrete mass,'which is'still at a higher temperature;"these cracks can be'

significant.' Problems of this type are primarily associated with massive

concrete'structures such' as dams rather than with LWR'concrete components. In addition, precautions to reduce hydration' effects include

using low or- moderate'heat' of hydration cients and "cool" materials as

well as partially replacing cement' idth fly ash''

Chemical reactions. Concrete cracking can also result from a number

of deleterious chemical reactions related generally to the aggregat'e

materials: alkali-aggregate reactions, 'cement-aggregate reactions, and

carbonate aggregate' reactions.,-

Expansive reactions between aggregates containing active silica and

alkalies derived from either cement" hydration, admixtures, or external

sources have caused many concrete'-structural failures in the past (late

1920s"-to early 1940s). The 'problem, which is generally confined to certain areas of the country, hwever,'has been significantly' reduced in'recent years through proper aggregate material selection,'use 'of low alkali

cements, and ddition'of pozzolanicmaterials.' The alkali-aggregate reaction'therefore should not be a problem for LWR concrete components,'

because these structures generally were all fabricated after 1960,'and

petrographic examination techniques were available to identify potentially reactive aggregates.

Highly siliceous aggregate materials in'Kansas, Nebraska, and Wyoming

areas have produced concrete deterioration'(map cracking) due to reaction

with alkalies' in 'cement.' This 'type' of distr'es's'should not be a, problem''

for LWR concrete components, becaus'e the' problem is regional,"and it can

be controlled by replacing 30% 6f tthe materials' with crushed limestone

aggregates.

-' Cert'ain dolomitic- limestone 'aggregates containing some clay and,

found in' only a' 'few geographical locations in the United'States and'

Canada 'react with alkal'ies to produce expansive reactions. This problem'

can be identified and controlled by diluting the reactive aggregate with

a less susceptible material;and -usingilow-alkali-content'cement.-'

4.1.2 Aggressive environments , ' -

Aggressive environments that could potentially lead to deterioration

of concrete include'weathering (i.e., freeze-thaw and wetting-drying),

leaching and efflorescence, and aggressive chemicals.
54

4.1.2.1 Weathering. Porous materials containing moisture are susceptible to damage under repeated cycles of freezing and thawing. Several different processes can contribute to the paste behavior during

freezing, including generation of hydraulic pressure due to ice formation, desorption of water from calcium silicate hydrate (C-S-H), and segregation of ice. Although a 9% volume increase occurs as water turns to

ice, which in turn will produce dilation in the microcracks, this is insufficient to produce all the dilation that occurs in concrete. The primary cause of dilation is-internal hydraulic pressure'generat ed by capillary water as it is being compressed during ice formation.7 When the hydraulic pressure exceeds the tensile strength of the cement paste, cracking occurs. If the concrete is partially dry or air-entrained, damage

will-not occur, because sufficient capillary space is available to prevent pressure buildup. Other factors leading to production of frost-.

resistant structures include: (1) selection of aggregates with adequate

durability* (certain rocks having fine pores and relatively high absorption combined with low permeability, such as cherts and shales, are susceptible to failure under freezing-thawing conditions); (2) use of low

water-cement ratio concretes properly handled, placed, and cured; and

(3) design of structures to minimize exposure to moisture and facilitate

drainage.

Alternate wetting and drying causes cycles of 'swellihg and'shrinkage. During periods of'increasing humidity, absorption of water on the

C-S-H surfaces creates a disjoining pressure. This pressure increases

with increasing thickness of absorped water (increased humidity) until it

reaches the point that it can exceed the van der Waal's attractions between adjacent particles, forcing them apart to create a dilation. Under

decreasing disjoining pressure (lower humidity), the particles are drawn

together by the van der Waal's forces- resulting in contraction. Disjoining pressure, is significant. only where the relative humidity is >50%.

4.1.2.2 Leaching and efflorescence. In structures containing areas

of poorly consolidated materials, cracks, or improperly treated construction joints, water may enter and pass through. As the water passes, someof the readily soluble calcium hydroxide and other solids are leached

out. With time this leaching can increase the concrete's porosity, which

in turn lowers its strength and increases its vulnerability to aggressive

chemicals. The rate of leaching is dependent on the amount of dissolved

salts' contained in the percolating water and on the water temperature

(calcium hydroxide is more soluble in cold water). Water flowing over

concrete's surface does not provide significant leaching.

Efflorescence is more of a surface phenomenon and consists.of deposited salts that have been leached from the concrete and are crystallized on subsequent evaporation of the water or on intersection with carbon dioxide in the atmosphere. Efflorescence is an aesthetic problem

*Certain aggregates [shales, clayey rocks, friable sandstones,

various cherts, and some micaeous material(s) that are readily cleavable

and structurally weak or are very absorptive and swell when moistened]

are subject to disintegration upon exposure to ordinary weathering

conditions. These materials can be identified through ANSI/ASTM C88

"Standard Test Method for Soundness'of Aggregates by Use of Sodium

Sulfate or Magnesium Sulfate."8
55

rather than a structural problem, but it is important in that it indicates that leaching is taking.place in the structure.

4.1.2.3 Aggressive chemicals. Concrete that is properly proportioned, placed,,and cured is relatively impervious to most waters, soils,

and atmospheres. Some chemical environments (in solution"form-above a'

minimum concentration),.however, can,cause-deterioration of even goodquality concrete. Because of the alkalinity of hydrated cement paste,'

alkaline materials usually do not ,attack it.' Acidic'materials, on the

other hand,' readily attack basic materials such as concrete through accelerated ,leaching of calcium hydroxide by~the hydrogen ion. Reference 9

lists various chemical agents and their effect on concrete as well as

commonly used protective treatments.-

Sulfates of sodium, potassium,.and magnesium present'in alkali soils

and waters havecaused deteriorationof concrete structures. The sulfates react chemically withthe hydrated lime cand hydrated calcium aluminate in cement paste to form calcium sulfate and calcium sulfoaluminate,

with considerable associated expansion and disruption of the concrete."

Sulfate resistance can be improved by the use of special sulfate-resisting

cements or admixtures, such'as pozzolans and blast-furnace'slag.

Sugar in solution is also very aggressive to concrete,'because it

dissolves-more than just calcium hydroxide; it attacks both'C-S-H and

calcium aluminate hydrates.

4.1.3 Embedment corrosion

Spalling and cracking of concrete can result 'from the corrosion of_

embedded' metals. Thecprim'ary embedded'material is reinforcing'steel,*

and the basic mechanism'is tensile forces created in the concrete through

formation of rust, which is an expansive reaction. Aluminum materials

embedded in concrete may cause the same destructive effects of corrosion

caused by (1) galvanic action between the aluminum and reinforcing steel,

(2) stray'electric-currents, and (3) alkalies in'the-concrete.6 Galvanic

corrosion of aluminum is accelerated'if chlorides and moisture are present. Lead and zinc behave somewhat like aluminum but to a lesser degree.

Copper and copper alloys have good resistance to corrosion unless chlorides are present.

4.1.4 Extreme environmental exposure

Extremeenvironmental-.conditions that could cause'deterioration'of

safety-related concrete-.components include'prolonged exposure to elevatedtemperatures and/or irradiation. -

4.1.'4.1 Elevated temperature -effects. Elevated temperature and

thermal .gradients are 'important to concrete'structures in that they affect the concrete's strength (ability to carry loads)'and stiffness'

(structural deformations and loads that develop at constraints)4' These

property variations result largely because of changes in the moisture'

content of the concrete constituents and progressive'deterioration'of''

*Corrosion of reinforcing steel is discussed in detail in Sect. 4.2.
56

the paste and aggregate (especially significant where thermal expansion

values for cement paste and-aggregate are markedly different). Other

factors of interest when a structure operates under elevated temperature

conditions are whether the component is under load (creep) or experiences

load cycling, the long-term'effects on strength and modulus of elasticity,

the concrete-rebar- bond strength, and the effectiveness of concrete

radiation shielding.

General behavior. Concrete made with portland-cement undergoes a

number of transformations when subjected to elevated temperatures.10 11

In addition to crystal transformations of the aggregate materials, a number of reactions occur to disintegrate the structure'of the matrix. At

low temperatures (<105 0C) these reactions' take'the'form of water expulsion. Dehydration of calcium hydroxide occurs when the temperature exceeds 400*C. Dissociation of calcium carbonate aggregates (if present)

is complete by 9OOC.* Above 1200C and up to 1300'C,'some components

of concrete begin to melt;' and some of the aggregates,'such as igneous

rocks (basalt), show degassing and expansion.- Above'1300 to 1400'C, concrete- exists in the form of a melt, with melting initiating in' the cement

paste matrix. Refractory concretes utilizing special-cements and aggregates are available for use in environments experiencing temperatures to

1800 to 2000'C, but they have not been used for fabrication of LWR components.

References 12-26 present results obtained from elevated temperature

testing of concrete. Figures 22 and 23 summarize some of the published

results on the residual compressive strength of concrete exposed to elevated temperatures for hot and cold testing, respectively.t Figure-24

summarizes' the effect of elevated temperature on concrete's residual

modulus of'elasticity for both open-hot and closed-cold conditions. Examples of the effect of. moderate elevated temperature exposure (T 180 0C)

on the stress-strain behavior of sealed and unsealed limestone concrete

specimens are presented in Figs. 25 and 26,. respectively.26

- Time-dependent response and thermal cycling. Time-dependent deformations (creep) at elevated temperature and thermal cycling can have an

*Aggregates containing'quartz undergo a crystalline transformation

from a-quartz (trigonal) to -quartz (hexagonal) between 500 and 650'C.

A substantial (5.7%) increase in volume accompanies this transformation. 1 1

tIn cold testing, the specimens are gradually heated to a specified

temperature, permitted to thermally stabilize at'that temperature for a

prescribed period of time, permitted to cool slowly-to ambient, and then

tested to determine mechanical properties. In hot testing, the specimens

are gradually heated to a specified temperature, permitted to thermally

stabilize at that temperature for a prescribed period, and then tested at

temperature to determine mechanical properties. During testing, specimens are maintained either in an open environment where-water vapor can

escape or in a closed environment where the moisture is contained. The

closed environment condition represents conditions for mass' concrete

where moisture does not have ready access to the atmosphere, and the open

environment represents conditions where the element is either vented or

has free atmospheric'communication.
57

ORNL-DWG 86-1724 ETD

100

x 80

* 0 * z

mu

w

*> 60

VjL)

IUIL

0

i U

-j 40

4(

20

0 L i 

0 200 400 600 800 1000 1200 1400 1600

TEMPERATURE (

0

F3

1 800

Fig. 22. Effect of 'temperature exposure on compressive strength

of concrete hot testing. -Source: . N." Freskakis et al.', "Strength

Properties of Concrete at Elevated Temperatures," Cip. *hg. NucZl. Power,

Vol. 1, ASCE National Conventioi,'Boston,'Mass.; April"1979. (References

noted in parentheses correspond with those cited in Ref. 25-in-Chap. '4.)

effect on the performance of concrete components with respect to increased

deformations (alignment) and potential strength loss, respectively.

Creep,* defined as an increase'in strain 'in a'structural member

with time due'to a sustained stre'ss," is important because it affects

'strains, deflections-and stress'distributions.t Figure 27 (Ref. 30) indicates' the evelompent of strain (creep) in a member with -age (time

since loading) and the effect that type-of aggregate can have on creep

magnitude' for concretes maintained at room temperature.' Like 'other

-*Creep of concrete in tension also occurs-and is of the same magnitude as -creep in compression.27 >-The follow~ing -discussion of creep will

pertain to creep under'compressive loadings. "

tBecause ,actual structures are generally under 'a multiaxial, stress

state, creep-of concrete under multiaxial loading is important., Information on creep of. concrete under multiaxial loadings, while at elevated

temperature, can be obtained.from Ref. 28, which presents data obtained

using a specially designed large'triaxial-torsion testing machine with

hygrothermal control (described in Ref. 29).
58

100

z

U

-J

0

uJ 

s0

60

40

ORNL WG81 1725ETD

NASSER 1104} a COMPRESSIVE . STRENGTH OF

CONCRETE AT ELEVATED

CAMPBELL-ALLEN (971 -TEMPERATURE-COLD TESTING-

\ \\ . By ~ONGISHI 11051 U_PPER BOUND DESIGN

RELATIONSHIP-TESTED

HOT OR COLD 11091

IANNANT( OLONERS 1107.

MALHOTRA 11021 IL B I

LOWER BOUND DESIGN ABRAMS (96\

RELATIONSHIP-TESTED COLD 1091

> \_ ~~~WIERIG I1951 (L.B.1 

FAULKNER (91

20

0

0 200 400 600 800 1000

TEMPERATURE (F)

1200 1400 1600 1800

-Fig. 23. Effect of temperature exposure on compressive strength of

concrete cold testing. -Source: G. N. Freskakis et al., "Strength Propertieseof Concrete at Elevated Temperatures," Civ. Eng. NucZ. Power,

Vol. l, ASCE National Convention, Boston, Mass., April 1979. (References

noted in parentheses correspond with those cited in Ref. 25 in Chap. 4.)

solids, creep of concrete increases.with temperature. Below 100'C, concrete creep-atmoderate stress levels originates in the cement paste,

probably because.of the.mutual approach of adjacent laminar particles of

cement gel, which is-facilitated by the presence of water in gaps between

the particles. 11 Another effect of temperature is the acceleration of

hydration-(aging) at.moderately elevated temperatures. At temperatures

above 105'C, dehydration occurs in a loaded concrete specimen, which

probably accelerates creep as shown in Fig. 28 (Ref. 19).

Thermal cycling, even at relatively low temperatures (650C), can

-have some deleterious-effects on concrete's mechanical properties. Results presented in.Refs. 31 and 32iindicate that the compressive, tensile, and bond strengths and the modulus of-elasticity are reduced and

that Poisson's ratio is increased. At higher temperatures (200 to

300°C), the first thermal cycle' causes the largest percentage of damage,

with the extent of damage markedly dependent on aggregate type and associated with ioss-of bond between the aggregate and matrix. The effect of

temperature cycling on a limestone aggregate concrete is' presented in

Fig. 29 (Ref. 31).
59

ORNL-OWG 81-1726 ETD

100

80

.0

IY

I-

-J

-J

0

U,

U-,

60

40

20

0

0 200 400 600 BO 1000 1200 1400 1600

TEMPERATURE I 0FI

1800

Fig. 24. Effect of temperature-exposure on modulus of elasticity of

concrete hot and cold testing. Source: G. N. Freskakis et al., "Strength

Properties of Concrete at Elevated Temperatures," Civ. Eng. NucZ. Pouer,

Vol. 1, ASCE National Convention, Boston, Mass., April 1979. (References

noted in parentheses correspond with those cited in Ref. 25 in Chap. 4.)

- ORNL-DWG 86-4570 ETD

c[N/mm2 j.

. J I 

Du

; C

W 50

w 40

UV 30

LU

cc 20

0

L)Io

I

I ' 1t200C .' -t I

a'/~~~~~~-7 

X ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-- - - X1 - / :50C~L

________ ________ ~CfIUnnI 0 2 3

- STRAIN ' 

Fig.<25. .Stress-strain-diagrams:of sealed limestone concrete specimens. Source: R. Kottas-et al., "Strength Characteristics of-Concrete-I

in the Temperature Range of 200 to.200 0

C," Paper Hl/2,-Sth Int'l. Conf.

on Structural Mechanics in Reactor Technology, Berlin, Aug. 13-17, 1979.
60

ORNL-DWG86-4571 ETD

aiN/mm2

]

60

M

t 40

w

w

3 20 0.

0

0

0 e(/o0o)

STRAIN

Fig. 26. Stress-strain diagrams of unsealed limestone concrete

specimens. Source: R. Kottas et al., "Strength Characteristics of Concrete in the Temperature Range of 20° to 200*C," Paper H1/2, 5th Int'l.

Conf. on Structural Mechanics in Reactor Technology, Berlin, Aug. 13-17,

1979.

ORNL-DWG 81-1720 ETD

1600

1200

If

e

, 800

w

0

400

0

10 28 90 1 2 5

DAYS YEARS

10 20 30

TIME SINCE LOADING (LOG SCALE)

Fig. 27. Creep of concretes with different aggregates. Source:

G. E. Troxell et al.; Long-Time Creep and Shrinkage Tests of Plain and

Reinforced Concrete, ASTM Proc. 48, pp. 1101-20, 1958.
61.

ORNL-DWG 86-4572 ETD

1000 a

750

cLO ~ - - TM NDRLADy

wj 500- IC

250- . o .

2 3 7 10 20. 50 100

TIME UNDER LOD. DaYs

Fig. 28. Creep of portland cement/porphyry concrete at various temperatures. Source: J. C.- Marechal, "Variations in the' Modulus of Elasticity and Poisson's' Ratio with Temperature," SP-34',;vols. 1-3, Paper

SP 34-27, Concrete-for-NucZear-Pressure-VesreZs, American-Concrete Institute, Detroit, 1972.

Long-term exposure-(aging). The design-lifetime of nuclear plants,

and thus of concrete components, is nominally 40 years. Over a plant's

operating' lifetime-certain 'concrete-components (i.e.,-the biological

shield-pedestal) 'iaaybe suj ected to moderately elevated temperatures,

which could affect the concrete's mechanical properties. Unfortunately

despite the potential significance of this' 'ffect,-only a limited number

of data have been identified relating the effects of long-term- elevated

temperature exposure (aging) to 'concrete properties.33 3b

Carette' t al.33 coiiducted an investigation to' det'ermine''the changes

in mechanical properties of a limestone aggregate concrete after exposures to temperatures up to 6000C for periods up to 8 months. For thermal exposure to 750C, compressive and splitting-tensile strengths after

8 months' exposure' were 98 and '94%, respectively, of'their reference

values.' However, on-exposure to 600'C for just'1 month, compressive-and

splitting-tensile strengths were only 23 and 38%, respectively, of their

reference values.. In companion'mixes,` where either fly ash :or blast

furnace'slag was used, 'no improvement in mechanical properties-'occurred

after' exposure to sustained high temperatures as a result of partial'replacement of the cement.
62

ORNL-DWG 86-4573 ETD

125

M 100 I-- u-c,

O z LUO Z

o : 75

-I- 6--

sZ 

' 50

ujLU

z

N

r---1, I ~I . -* -

1 0

.

LUWz

12 5.T - r - -

I *% I 

100 -. F_ _

1-4-4-1-8-

-- - LIMESTONE CONCRETE

(FINESANDCOARSE -

(LL3)

- SAND, LIMESTONE - CONCRETE LN3)

1 I I 1I

Cx - '- 

Cz o

LU18

0IF 50

O M

z

D 25

650C

-o COMPRESSIVE STRENGTH -

o TENSILE STRENGTH

x MODULUS OF ELASTICITY

I t I I

0 5 10 15

NO. OF CYCLES

CEMENT 3

20 C 5 10 15

NO. OF CYCLES

CEMENT 2

20

100

OL Z

o z

LU LU

z 

LU LU

IC<

z

10 20 0 10

NO. OF CYCLES NO. OF CYCLES

CEMENT 3 200 0 C CEMENT 2

20

100

I0 75

L, UJ.

-

Z 0.5

UJ U 1

-2

CL25

z

100

M

I0z

Z LU

z

75

50

25

0 10 20 0 10 20

NO. OF CYCLES NO. OF CYCLES

CEMENT3 300 0C CEMENT2

- - - LIMESTONE CONCRETE (FINES AND COARSE) (LL3) 0 COMPRESSIVE STRENGTH

0 TENSILE STRENGTH - SAND, LIMESTONE CONCRETE (LN3) x MODULUS OF ELASTICITY

Fig. 29. Effects of temperature cycles on limestone concretes.

Source: D. Caipbell-Allen and P. M. Desai, "The' Influence of Aggregate

on the Behavior of Concrete at Elevated Temperature," NucZ. Ehg. Des.

6(1) (1967).
63

Mears34 investigated the effect of long-term exposure (up to

13 years) at moderately elevated temperature (650C) on the mechanical

properties of a limestone aggregate concrete. These tests' were somewhat

unusual in that the specimens were first subjected to a simulated

temperature-vs-time cement hydration cycle. Also, because the concrete

mix was being evaluated for an application that experienced exposure to

sulfate-bearing groundwater at elevated temperatures (65 0 C), both ordinary and sulfate-resistant portland cements were investigated. Specimens, after being subjected to the simulatedcement hydration cycle, were

stored either in water at 19'C (control specimens) or in a sodium sulfate

solution of 2000 ppm at 65C.- Frequently during the test program, the'

sodium sulfate solution was changed, which required cooling to room temperature; the specimens were therefore also subjected to thermal cycling.

Results of the study'indicated that there was no evidence of long-term

degradation in compressive strength for any of the.concrete-mixes and

heat treatments utilized and that-for a given compressive strength, the'

dynamic modulus of elasticity was -lower for the concrete that had been

heated.

A five-year testing program was conducted to determine the effects'

of long-term exposure to elevated temperature on the mechanical properties of concrete used in constructing the radioactive underground storage

tanks at Hanford Engineering Development Laboratory (HEDL).35 Tests were

conducted using specimens fabricated from the same mix proportions and

materials specified for the concrete used to fabricate the tanks (20.7-

and 31.0-MPa design compressive strengths). Concrete strength, modulus

of elasticity, -and Poisson's-ratio values were determined from specimens

subjected to either'121','177,-or 2320 C for periods-of up-to 33-months.

The effect of thermal cycling was also investigated.. Results showed that

the compressive strengths in general'tended to decrease with increasing

temperature and also with length of exposure;-however, with the exception

of the cylinders exposed to 232C, all compressive strength results obtained after a 900-d exposure exceeded design values.' Splitting-tensile

strength results also decreased -somewhat with increasing temperature and

length of exposure. Modulus of elasticity was affected most significantly by. the elevated temperature exposure; after 920 d of heating at -

232°C, it had a value of only-30% - the value obtained from an unheated,

control specimen. Poisson's ratio, although exhibiting somewhat erratic

values, was relatively unaffected by either th&'magnitude or the length

of'elevated temperature exposure. Thermal cycling (18 cycles) to 177°C

produced moderate reductions in compressive strength (5 to 20%), significant reductions in modulus (30 to 50%), and slight reductions in

Poisson's ratio (0 to 20%).- Time-dependent. (creep) and,-thermal property

data were also obtained from the concrete mixes.

Associated with the laboratory investigation described in the previous paragraph was a study to confirm the laboratory results by testing

samples-removed-from- the underground-storage-tanks-and process buildings

at HEDL.36 Cores 76 mm in diameter were obtained over the length of the

haunch, wall, and footing of a single-shell tank that was built in 1953;

contained' waste''for about'B years; reached temperatures in the range of

127'to i380 C; and experienceda radiation field "of '0.10 to 0.13 C/kg/h

(400 to 500'R/h). Although considerable'scatter was obtained'from the
64

data because of different concrete pours and different environmental exposures, after about 29 years of exposure, only one data point fell below

the 20.7-MPa design compressive strength. Figure 30 presents compressive

strength results obtained from these tests as well as those obtained from

tests on concretes from other structures and compares them to predicted

values obtained from laboratory work.

Concrete-reinforcing steel bond strength. Only limited data are

available on the effect of elevated temperatures on the bond strength between concrete and steel reinforcement. agami37 - in testing specimens

fabricated from river gravel concretes containing embedded plain, round,

steel bars - found: that the residual bond stress after subjecting the

specimens to 300%C for 90 dand then cooling to room temperature was only

-50% the value beforeheating. Milovanov and Salmanov38 demonstrated the

importance of reinforcement type when they showed that ribbed bars experienced a loss of bond strength-only above 400C but that smooth bars

lost strength after only a small temperature increase. Results presented

in Refs. 39 and 40 indicate that for exposure temperatures <150%C the

loss in bond strength between concrete and steel reinforcement is small

(<15%). -

-Radiation shielding'effectiveness. Portland cement concrete possesses many of the physical qualities of an ideal radiation shield. It

ORNL-DWG 86-4574 ETD

It, 6M.

LABORATORY RESULTS. 700 F

_. -. ---- LABORATORY RESULTS. 260°F

. PUREX CORE 72?F) FROM SAMPLE GALLERY SHIELDING WALL * -

100- * PUREX CORE'(200 0 F) FROM N & M CELL .

o 10.000 A TANK FARM CORE (T£STED AT 720F) - (DOME) 0 a z

u. A TANK FARM CORE (TESTED AT260F) -(DOME) . o

PUREX CORE 720 F) FROM N & M CELL 0oo°° -

TANK FARM COREAT72F(HAUNCH&WALL) o

m 8.000 / ° * 6

aa 

C~~~~~~~~~~~~~~~~~~~~~ao /' * , 0t . . .. 

. 6.000@ 2 *** 

400 800 1200 25 YEARS

T~~~~ . . - ..-.

Fig. 30. Laboratory vs actual-sample compressive strength data.

Sour'ce:, M. P. Gillen et al.; strength and Elastic P,opertiesof Concrete

Exposed to Long-Moder'ate emperatures and High Radiation Fields, Rib-RESA-55 P. Rockwell Hanford Operations, Richland, Wash., 1984.
65

is a polyphase material consisting of particles of'aggregate contained in

a matrix of portland-cement-paste. Gamma'rays are absorbed by-the highdensity aggregate materials, and eutro'ns'are'atteniuated by'hydrogen

atoms in the cement paste. The effectiveness of concrete as a-shield,

however,'may be reduced under-service conditions (elevated temperature)

as drying reduces the hydrogen content-or cracking occurs.

Results'of elevated 'temperature exposure 'on shielding of heavyweight aggregate'(iron limonitetand magnetite limonite) concretes are

presented in Fig. 31 (Ref.'41). Significant changes in'attenuation effectiveness were found as -the concrete was heated to 100 and 175%C, with

little additional change effected'in heating to 3200C.' Despite'the loss

of'neutron nd'gamma attenuation'efficiency' with increasing temperature,

it was concluded'that the concrete would serveas a satisfactory shield

material. 'If increasing efficiency'were required at'higher temperatures,

it could be'accounted'fo'r in the design.- ' '

* ~~~~~~ ~- - - * .- 

ORNL-DWG 86-4575 ETD

10~ 105 '' '......-I1,.

*- . 10 000C

'(N E ' - - d 20C.

U- ' ;'- -

200C -

X 3000C

-3

z 

0

I-

-. ~z AS-CURED

A2 -

101 -

1 0~~~~~~-

-- ~0 10 20 30 40- 50

INCHES. 

Fig.A31. ''Thermal' eutrorndistribution'in ordiiiary concrete as a

function of t'emperatur&.' Source: E. G.-'Peterson',` 1Shielding 'Properties

of Ordinary Concrete as a FTnction of Temperature,'HW-65572, Hanford

Atomic'Products Operation,'-Richland, Wash., Aug. 2, 1960.
66

Shielding effectiveness of concrete is also reduced if through

cracks develop., Reference 42 investigated the effect of gamma rays

through a concrete shield containing straight and crooked cracks. In the

immediate vicinity of the concrete surface, leakage of gamma rays through

a slit contributed significantly to y-dose rate but diminished rapidly

with distance from the surface as a result of shield thickness and scattering effects. Reference 43 investigated the shielding effectiveness of

cracked concrete and developed formulas to define the resulting effects.

Guidelines developed for compensation for cracking concluded that it

might be economically advantageous to allow a concrete shield to crack

and then shield the resulting irradiation by other means.

4.1.4.2 Irradiation effects. Concrete has traditionally been used

as a shielding material because it attenuates. radiation with reasonable

thickness requirements, has sufficient mechanical strength, can be constructed in virtually any size and shape at reasonable cost, and requires

minimal maintenance. Irradiation, however, in the form of either fast

and thermal neutrons emitted by the reactor core or gamma rays, produced

as a result of capture of neutrons by members (particularly steel) in

contact with the -concrete- can affect the concrete. The fast neutrons are

mainly responsible for the considerable growth, caused by atomic displacements, that has been measured in the aggregate. Gamma rays produce

radiolysis of water in the cement paste, which can affect concrete's

creep and shrinkage behavior to a limited extent and also result in evolution of gas.

Operation of a reactor over its 30- to 40-year life expectancy may

subject the concrete to considerable fast and thermal neutron fluxes.

Reference 44 estimates the following values for maximum radiation to

which the prestressed concrete reactor vessel of a high-temperature gascooled reactor could be subjected after 30 years of service:

thermal neutrons: 6 x 1019 neutrons/cm2,

fast neutrons: 2 to 3 x 1018 neutrons/cm2,

gamma radiation: 109 Gy (1011 rad).

For a 1250-MW(e) pressurized-water reactor, Ref. 45 estimates the integrated flux exposure to the inner surface of the biological shield as 1019

fast neutrons/cm2 after 40 years of service. Section III, Division 2, of

the ASME Boiler and Pressure Vessel Code 46 gives a radiation exposure

level allowable to 10 x 1020 neutrons/cml. The British Code for prestressed concrete pressure vessels47 states that the maximum permissible

neutron dose is controlled by the effects of irradiation on the concrete

properties, and the effects are considered to be insignificant for exposure levels up to 0.5 x 1018 neutrons/cm2. Note, however, that these

criteria are based on a very limited number of data and that quantifying

the extent to which irradiation will change the properties of concrete is

impossible because sti-ch quantification is dependent on many factors, such

as variation of material properties, material state of testing, neutron

energy spectrum, and neutron dose-rate.

Several reports have been written on the effects of irradiation on

concrete propertie y448-75 The apparent availability ofdata on irradiation effects on concrete properties is, however, misleading because

of technical and experimental difficulties in conducting meaningful
67

tests. In addition, available data are generally not comparable because

(1) different materials were used, (2) mix proportions varied, (3) specimen size-was inconsistent, (4) temperatures varied, and (5) both cooling

and drying conditions were-used. Reference 44 presents an excellent summary of experimental data that are available on irradiation effects on

concrete properties. Twelve conclusions can be drawn from these data.

(1) For some concretes, neutron-radiation of >1 x 1019 neutrons/cm2 may

cause some reduction in compressive strength (Fig. 32) and tensile -

strength (Fig. 33). (2) The decrease of tensile strength due to neutron

radiation is more- pronounced than the decrease of compressive strength.

(3) Resistance of concrete to neutron radiation apparently depends on the

type of neutrons (slow or fast) involved, but the effect'is not clarified. (4) Resistance of concrete to-neutron radiation depends on mix

proportions, type of cement, and type of aggregate (Fig. 34). (5)'The

effect of gamma radiation on concrete's mechanical properties requires

clarification. (6) The deterioration of concrete properties associated

with a temperature rise resulting from irradiation is relatively minor.

(7) Coefficients of thermal expansion and conductivity of irradiated concrete differ little-from those that-would result from temperature-exposed

concrete.. (8) The modulus of elasticity~ of concrete when exposed to'neutron irradiation decreases with increasing neutron fluence (Fig. 35).

(9) Creep of concrete is not affected by low-level radiation exposure,

but for high levels of exposure creep is likely to increase with exposure

because of the effects of irradiation on the concrete's tensile and compressive strengths. (10) For some concretes, neutron radiation with a

fluence of >1 x 1019 neutrons/cm2 can cause a marked increase in volume.

(11) In general, concrete's'irradiation resistance increases, as the, irradiation resistance of aggregate increases. (12) Irradiation has little

effect on shielding properties of concrete -bey6nd the effect of moisture

loss due to a temperature increase.

Although detailed information is very limited, Ref. 66 indicates

the effect of extended periods of irradiation on concrete properties. In

the study, concrete was removed from the 2.13-m-thick graphite reactor

shield (Fig. 36) at Oak Ridge National Laboratory after being in place

for 12 years.. To obtain a complete picture of the ,conditions towhich

the shield had been-subjected, temperature gradients (19- to-.40 0C), gamma

ray [8.1 x 10-8 (8.1 x 10-1) to 1.9 x 10-1 J/g-h (1.9 x 106 erg/g-h)] and

fast-neutron [undetectable to-1.78 x 10-3 J/gh (1.78 x 104 erg/g-h)]

dose rates, and thermal-neutron fluxes (1.88 x 102 to 4.47 x 1010 neutrons~ca 2.s)':weredetermined.*.;Analysis.of a 117-mm-diam core sample

through the'shield showed that the'-chemical properties and density of -the

shield had not changed 'appreciably since a similar investigation done

8 years'earlier; however,-the compressive strength at the reflectorshield interface had dropped-as much'as 40% (16.9 to 10.1-MPa), while..

near the back-of'the shield (thickness --2.0 m) the change was negligible

(11.4 to-11.1 MPa). Damage-to the concrete by irradiation was felt to-be

less than that caused by related temperature effects.

*Highest values were at the reflector-shield interface (thickness =

0 m) and lowest values at the back of shield (thickness = 2.13 m).
I

68

ORNL-DWG 81-1738A ETD

1.4

1.2

1.0

0.8

0

_u

I I. I I I1 1 1 1I 

SLOW NEUTRONS: FAST NEt

ALEXANDER (151) 0 ELLEUC

- A A PRICE (165) A HOUBEN

V STOCES

AG-ANOIN F1

0_ i ,~~~~~~ DUBRO\ - S~~~~~~~~~~~~~~

JTRONS: -

H (158)

I (161)-

(168)

BRVATIO -

( 152)

fSKIJ 157)-

0.6

0.4

0.2

0

ta) I I.I I'I* I I I I . I I I

1.4

1.2

1.0'

-I- 1;

_ 0.8

_U

0.6

0.4

0.2

.

0 I_

10i

2 5 1019 2 5 10 2 5 021 2 5 102

FLUENCE OF NEUTRON RADIATION (neutrons/cm2 )

Fig. 32. Compressive strength of concrete exposed to neutron

radiation relative to untreated concrete: thermal effects on strength

(a) not included, (b) included. Source: H. K. Hilsdorf et al., "The Effects of Nuclear Radiation on the Mechanical Properties of Concrete,"

Douglas McHenry International Symposium on Concrete and Concrete Structures, Publication SP-55, American Concrete Institute, Detroit, 1978.

(References noted in parentheses correspond to those cited in Ref. 25 in

Chap. 4.)
69

ORNL-DWG 81-1739A ETD

I I I I I I

I I :

1.4

1.2

O ELLEUCH

A GRAY -

A GRAY

- . 1.1* BATTEN-

(158)

(159)

(159)

(152)

6 I I I

FAST NEUTRONS O.n.)

LIMESTONE AGGREGATE f.n.)

FLINT AGGREGATE (f.n.)

NO INFORMATION

1.0 -

a 0.8 1_

0.6 -

OA _

0.2 _

(a) i

0 I I I I I . I , . I ' I 

1.4

1.2

1.0

-.

0.8

I* I- I I I I1, . .. 

IbJ~~~~~~ 0 ---- I

0.6

OA

0.2

0

1018 2 5 -0

19 l

o25 0 2 5 o21 2

FLUENCE OF NEUTRON RADIATION 2utronslcm

Fig. 33. Tensile strength of concrete exposed 'to neutron radiation

relative to unt ated cncrete,. thermal effects onstrength (a) not included, (b) included. Source: *H. K. Hilsdorf et al., "The Effects of

Nuclear Radiation on ,the Mechanical Properties of Concrete," Douglas

McHenry.International Symposium on Concrete and 'Concrete Structures, Publication SP-S5, American Concrete Institute,Detroit, 1978.- (References

noted in parentheses correspord,to those citedin Ref 25 in Chap. 4.)
70

ORNL-OWG 81-1740A ETD

5

4

w

3

1

0

-I

0 1 2 3 4 x 1019

FLUENCE OF NEUTRON RADIATION. nlcm2

Fig. 34. Effects of fast neutron exposure on volume change of flint

aggregate and limestone aggregate concretes. Source: H. K. Hilsdorf

et al., "The Effects of Nuclear Radiation on the Mechanical Properties of

Concrete," Douglas MHenry International Symposiumon Concrete and Concrete Structures, Publication SP-55, American Concrete Institute,

Detroit, 1978. (References noted in parentheses correspond to those

cited in Ref. 25 in Chap. 4.)

& i ORNL-DWGSI-1742A ETD

* ALEXANDER (151) S.N.

1,4 - _ 0 OUBROVSKU 11571 NO INFORMATION

° GRAY 11591 F.N.

HOUSEN (161I F.N.

1.2 - 0 v. d. SCHAAF (1661 F.N.

V STOCES (168) F.N.

8

* I wU

1.0- 

0.2 - - - 0 -

.2 -

1018 2, 5 1019, 5- 10202 5 1021 2

FLUENCEOF NEUTRON RADIATION, nIcm2

Fig. 35. Modulus of elasticity of concrete exposed to neutron

radiation relative to untreated concrete: thermal effects on'modulus not

included. Source: H. K. Hilsdorf et al., "The Effects of Nuclear Radiation on the Mechanical Properties of Concrete," Douglas McHenry International Symposium on Concrete and Concrete Structures, Publication

SP-55, American Concrete Institute, Detroit, 1978. (References noted in

parentheses correspond to those cited in Ref. 25 in Chap. 4.)
71

ORNL-DWG 86-4576 ETD

/g TAR LAYERS j |r GRAPTE

BARYTES-HAYDITE:.'

. .,. : ^ CONCRETE ;:. e o o e VY

; ¢' *~~~~~~~,, I.. ,{s ljoOX

a~~~~~~~~~~

itt W~~~* . Lft

114 in. 1/4 in. k.3 c

Fig. 36. Cross section'of ORNL graphite reactor shield.' Source:

T. V. Blosser et al., A Study of the Nuclear 'and Physical Properties of

the ORNL Graphite Reactor Shield, ORNL-2195, Aug. 25, 1958.

4.2 Concrete Reiforcing Steel egradation

Mild steel reinforcing bars are'provided to'control the extent.of

cracking and-'the width'of cracks- at operating temperatures, resist ,tensile stresses 'and compressive stresses for elastic design, and provide

structural reinforcementwhere required by'limit condition design.procedures. 4 6 176 Potential causes of degradation of the reinforcing steel are

corrosion, elevated temperature exposure, and irradiation. '''.

4.2.1 Corrosion ' - : -

When portland cement hydrates, the silicates-react.with water to

produce calcium silicate-hydrates and calcium'.hydroxide. The .high alkalinity of this chemical.environment normally protects embedded steel

because of.the formation of a rotective oxide film (gamma Fe2 O3)-on the

steel. Passivity of thi's'p'rotective'film, however', ;can be destroyed by

penetration of aggressive ions or a reduction.'in' the' pH to 411,"which can

be caused by leaching of alkline substances by water or by reaction with

carbon dioxide or acidic materials. 7 7 Carbonation," which is discussed

earlier, is primarily a surface effect of insignificance unless the concrete is of poor quality or the rebar has very shallow cover. Leaching
t

72

by flowing water or reaction with acidic materials is felt to be only a

remote possibility for LWR concrete components. Therefore, the most

likely cause of steel reinforcement corrosion is related to chloride

ions.

Possible sources of chloride ions include aggregates containing

chlorides, saline water used as' mix water, calcium chloride accelerators,

cements containing small amounts: of chlorides, and the environment.* For

steel corrosion to occur, four essential elements of an electrochemical

cell must be present ('1t) anode'(point of electron release and where

ions go into solution), (2)-conductor (rebar), (3) cathode (electrons

consumed in presence of oxygen and moisture), and (4) electrolyte (moist

concrete). When the mietallic- iron of the steel reinforcement is converted to rust [Fe(OH)3], a volume increase of 600 to 700% occurs, which

causes cracking and spalling-of' concrete where its tensile capacity is

exceeded. The extent of corrosion is somewhat dependent on the orientation and geometry of the crackt:and on time.

Most research reports and'the ASME Code provisions deal with the

width of the crack at the concrete surface, which is not in any way

uniquely related to the crack-,width at the rebar. Width of a crack at

the level of the rebar is related to the crack origin, amount of concrete

cover, steel stress, bar'diameter-, reinforcement ratio, arrangement of

bars, and depth of the tensile zone. When a crack is transverse to the

rebar, localized corrosion occurs only over'about three bar diameters.

When the crack is longitudinal and coincides with the rebar, passivity is

lost'at many locations, and corrosion can proceed unchecked. eeby79 relates corrosion to time by noting that

to + t > design life of structure

where to is the initiation phase (time from construction to passivity of

rebar is destroyed) and t is the active phase during which corrosion

occurs. If to and t can be established, the life of the structure can

be determined; however;,inreality this is difficult because to depends

on several unknowns (environment, concrete permeability, whether cracking

has-occurred, and cover) as does t, which also must account for rate effects-and defining'an acceptable level of corrosion.

*Cracks in concrete accelerate the'onset of corrosion that results

from chloride ion penetration; however, the corrosion is confined to the

point of intersection with the reinforcement. Some analysts feel that

because chloride ions eventually can penetrate uncracked concrete to initiate more widespread corrosion, little difference exists between the

amount of corrosion in-cracked and uncracked concrete.78.

fReference 78 presents information on'tolerable crack widths in

reinforced oncrete'structures for different exposure conditions.

Building codes protect reinforcement from'corrosion by (1) specifying

minimum cover, (2) establishing minimum concrete quality, and (3)

limiting crack widths.
73'

4.2.2 Elevated temperature effects

The properties of reinforcing steel used in design are generally a

function' of the yield stress, which is affected by exposure to elevated

temperature. The ,ield strength of ANSI/ASTM A36 structural steel is

relatively unaffected for temperature exposures <930 C (Ref. 80). Data

for German reinforcing steels (Fig. 37) indicate that for temperatures up

to'-200%C the yield strength is reduced 10% and that at 5000 C it falls

to -50% of its reference value, with hot rolled steels performing better

than''cold twisted or cold drawn steels.10 The modulus of elasticity exhibits a similar reduction pattern with increasing temperature.

ORNL-DWG 86-4577 ETD

~140 o e 1) HOT ROLLED STEEL . BSt 220/340

2) HOT ROLLED STEEL

120 - 'St 420I500

- -. 3) COLD TWISTED STEEL

CIA BSt 420/500

Z 100 41 COLD DRAWN STEEL

BSt 420/500

e . .. . ~ . ~~~ ..

i

Co 4

, ~ ~ ~ ~ ~ _ 2

-J

-. .. 2

43

0 200 .400 600 800 1000 1200

TEMPERATURE IN CC

Fig.~37. Effect of temperature on yield strengt-h (0.2%) ofjfour

types of German reinforcing steelI. Source: U. Schneider et al., "Effect

of Temperature on%,Steel and Concrete for PCRV's," NucZ. Eng. Dec. .67,

245-58 (1981).

4'.2.3 'Irradiation effects

Neutron irradiation-produces changes in the mechanical properties of

structural'steel, for example, an increase in the materials yield

strength and a rise in the ductile/brittle transition temperature.47

These changes are shown in Figs. 38 and 39, which present the effects of

irradiationf'on the stress-strain curve for a mild steel and Charpy V-notch

energy and temperature curves for unirradiated and irradiated mild steel,

respectively.8 1
ORNL-DWG 86-4578 ETD

50

45

40

35

- 30

d

in

z

o 25

I-_

15

10

5

1 0

ORNL-DWG 86-4579 ETD

m

cJ IIL

a

wr

2:

w

10 20 30

STRAIN: 

40 60 60 100 . 140 180 220

TEMPERATURE: OC

Fig. 38. Stress-strain curves for unirradiated and irradiated mild steel. Source: A. Cowan

and R. W. Nichols, "Effect of Irradiation on Steels

Used in Pressure Vessels," Group D Paper 20, Prestressed Concrete Pressure Vessels, The Institute

of Civil Engineering, London, 1968.

Fig. 39. Charpy V-notch energy/temperature

curves for unirradiated and irradiated mild steel.

Source: A. Cowan and R. W. Nichols, "Effect of

Irradiation on Steels Used in Pressure Vessels,"

Group D Paper 20, Prestressed Concrete Pressure

Vessels, The Institute of Civil Engineering,

London, 1968.
75

4.3 Concrete Prestressing Steel Degradation

A osttensioned prestressing system consists'of'a-prestressing tendon in combination with methods of stressing'and'anchoring the tendon.to

harden concrete.. To attain satisfactory pe'rformance,'prestressing .systems:are designed to have (1) consistently high'strength-and strain.at

failure,' (2);serviceability. throughoutt their lifetime, (3) reliable and

safe.prestressing procedures, and (4) the ability to be retensioned and

replaced (nongrouted'systems).82 Prestressing-systems may be grouped

into three major categories, depending-on'the type of tendon used: wire,

strand, or'-bar. In the United 'States the-8.9-MN systems, which are approved for use in containments, include'(1) BBRV (wire), '(2)'VSL.

(strand), and (3) Stressteel S/H (strand). Potential.degradation modes

for-these prestressing systems'include corrosion, elevated temperature

exposure,-and-irradiation. '

4.3.1 Corrosion

Corrosion may behighly localized or uniform. Most 'prestressing

corrosion-related failures have been-the.result of localized attack

produced by pitting, stress corrosion, hydrogen embrittlement, or combinations of these. Pitting is an electrochemical process that results in

local penetrations into the tendon to reduce the cross section to the

point where it is incapable of supporting its load. Stress corrosion

cracking results in.brittlefracture of.a normally. ductile metal or alloy

under.stress (tensile or'residual) in specific corrosive environments.

Hydrogen embrittlement, frequently associated with hydrogen sulfide, occurs when hydrogen atoms enter the metal lattice and significantly reduce

its ductility. Protection of:the-prestressing systems is provided by

filling the ducts containingthe posttensioned tendons either with microcrystalline waxes (petrolatums) compounded'using organic corrosion inhibitors (nongrouted tendons) or with portland cement grout (grouted tendons). Regulatory requirements for'inspection'and replacement have made

nongrouted posttensioned steel tendons the dominant prestressing system

used in containments.

Reviews 83 -

85 of the performance of prestressing.tendons contained in

both nuclear power plant and conventional civil engineering structures

indicate that corrosion-related incidents are'extremely limited (see

Chap. 3). The evolution"of'corrosion inhibitors and'the use of organic

petrolatum-based compounds designed especially' for corrosion protection

of prestressing materials have significantly reduced corrosion of prestressing materials. The few incidences of corrosion'that were identified generally occurred early in the'use of prestressed concrete for containment structures and-either resuited-from--the use of off-the-shelf

corrosion inhibitors that had not been specially formulated for prestressing materials or were!the'result of poor construction practices.

The problems were subsequently identified and corrected during the construction phase, the initial structural integrity test, or subsequent inservice inspections.
76

4.3.2 Elevated temperature:effects

The effect of elevated temperature on all heat-treated and drawn

wires can be significant, and'on cooling they do not regain their initial

strength because the heating destroys the crystal transformations achieved

by-the heat process. Short-term heating, on the Iorder of 3 -to 5 min,

even to temperatures as-hi h'as 400 C, however, may.not do any harm.86

Results of a Belgian study involving 30 types-of prestressing-steels

indicate' that' thermal exposures up to 200C-do not significantly reduce

(<10%) the tensile strength of, prestrdssing wires or strandsl[see Fig. 40

(Ref. 10)]. Stress-strain curves for ANSI/ASTM A 421 steel (stressrelieved wire for prestressed concrete) as a function of temperature are

presented in Fig.'41 (Ref. 80).- - -

Elevated temperature exposures also affect the-relaxation and creep

properties of prestressing tendons. An indication of the effect of moderately elevated temperatures 200C T 100"C) on the relaxation of a

low-relaxation strand with data extrapolated to 50 years is presented in

Fig. 42 (Ref. 87). Reference 88 indicates that losses in a 15.2-mm-diam

strand initially stressed to 75% guaranteed ultimate tensile strength at

40C will be 5 to 6.4% after 30 years. Relaxation losses of tendons composed of stress-relieved wires have relaxation losses of about the same

ORNL-DWG 86-4580 ETD

140 .

o . 30 TYPES OF STEELS

.. . - WIRES: 05to7mm

* cJ .120 - STRANDS: 05to12mm

o . - STRENGTH: 1500 to 2200

8 N/mm2 at 200C 73100

-0 2 .,

z I 70% OF ALL DATA

w - - . EPRTREIN°

40

C,20 _

0 -

-.0 200 400 600 800 1000 .1200

TEMPERATURE IN 0 C

Fig. 40. Effect of temperature on ultimate strength of 30 different

types of prestressing steel. Source: U. Schneider et al., "Effect of

Temperature on Steel and Concrete for PCRV's," NucZ. g4. Des. 67, 245-58

(1981).
77

ORNL-DWG86-4581 ETD

250

A4210200

STE .40TRI0.i./n

0

: . . ..

0 . 1100 F1220 F0 0.01 0.02 0.03 0.04. 0.05 0.06 0.07 0.08 0.09 0.10 0.11 0.12

STRAIN in.in. ..

Fig. 41. Stress-strain curves for ANSI/AST A 421 steel at various,

temperatures. Source: T. Z. Harmathy and W. W. Stanzack, Elevated TeR-.

perature Tensile and Creep Properties of Some Structural and Prestressing

Steels, ASTM STP 464in Fire:Test Performance, 1970.

OR NL-DWG 86-4582 ETD

100

014

.2

10 ~ 10- ,00 10,000 100.000 30 50

Time (Hours) - Yes

Fig. 42.' Relaxation vs-time-6crve for low-relaxation-strand.

stressed to 70% GUTS 'and held at various temperatures.' Source: J. R.

Libby, Modern Prestressed Concrete, Design Principles and Construction

Methods, Von Nostrand Reinhold Co., New York, 1971.
78

magnitude as stress-relieved strand, but relaxation of a strand is

greater than that of its straight constituent wire because of the combined stress relaxation in the helical wires.89 Creep (length change

under constant stress) of stress-relieved wire is negligible up to 50%

its tensile strength. Also, the creep effect in steel varies with its

chemical composition as'well as with mechanical and thermal treatment applied during the manufacturing process.

4.3.3 Irradiation effects

Irradiation of steel affects its mechanical properties because atoms

are displaced from their normal sites by high-energy neutrons to form interstitials and vacancies. These defects can grow together and effectively both strengthen the steel and reduce its ductility; or, at higher

temperatures, they can recombine and annihilate each other and, for a

given neutron dose, reduce the irradiation damage.81 Results obtained

from studies81 in-which 2.5-mm-diam prestressing wires were stressed to

70% of their tensile strength and irradiated to a total dose of 4 x 1016

neutrons/cm2 (flux of 2 x 1010 neutrons.cm2 -s) showed that for exposures

up to this level the relaxation behavior of irradiated and unirradiated

materials was similar. Because these flux levels are higher than the

level likely to be experienced in a LWR containment, it does not appear

that irradiation of prestressing will have a harmful effect on the containment.*

4.4 Anchorage Embedment Degradation

Anchorage to concrete is required for heavy machinery, structural

members, piping, ductwork, cable trays, towers, and many other types of

structures. An anchorage might have to meet certain requirements for

ease of installation, load capacity, susceptibility to vibration, preload

retention, temperature range, corrosion resistance, postinstallation or

preinstallation, and ease of inspection and stiffness.91 In meeting its

function, loads'that the anchor must transfer to the concrete vary over

a wide combination of tension, bending, shear, and compression. Examples

of types of anchors available include embedded bolts (A-307, A-325, or

A-490), grouted bolts, embedded studs, self-drill expansion anchors, and

wedge anchors. Several potential factors related to failure or degradation of the anchorage systems.include design detail errors, installation

errors (improper embedment depth or insufficient lateral cover, improper

torque), material defects (low anchor or concrete strengths), shear or

shear-tension interaction, slip, and preload relaxation.91'92 Aging

*Irradiation of-corrosion inhibitors such as used in PCCs of LWR

plants indicates that there are no changes outside of the specification

ranges in physical and chemical properties of the corrosion inhibitors

when irradiated to-105 Gy (1 x-107 rad). This exceeds the gamma radiation-level expected during the 40-year life of a nuclear power plant.90
79

effects that could impair the ability of an anchorage to meet its performance requirements would be primarily those that result in a deterioration of concrete properties, because if a failure did occur, it would

most likely initiate in the concrete.

References

1. B. Mather, "Concrete Need Not' Deteriorate," J. Am. Concr. Inst.

1(9), 32-37 (September 1979)..

2. ACI Committee' 116, Cement and Concrete Terminology, ACI SP-19,

Detroit, 1967.

3. S. Mindess and J. F. Young, Concrete, Prentice-Hall, Englewood

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4. J. Rpke, Concrete Problems, Causes and Cures, McGraw-Hill Book Co.,

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5. W. H. Price, "Control of Cracking During Construction,", Concr. Int.

4(1), 40-43 (anuary-1982).-

6. G. E. Troxell, H. E. Davis, and J. W. Kelly, Composition and Properties of Concrete, 2n,! ed., McGraw-Hill Book Co., New York, 1968.

7. T. C. Powers, "Resistance of Concrete to'Frost at Early Ages,":pp.

1-47 in Proceedinge RILEM Symposium on 'Winter Concreting, Session C,

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8. "Standard Test Method for Soundness of Aggregates by Use of Sodium

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for PCRVs," NuCZ. Eng. Des. 67, 245-58 (1981).-;

11. Z. P. Bazant et al., ,Normal and Refractory Concretes for LMFBRApplications - VoZ. 1: Review of Literature on High-Temperature 'Behavior of Portland Cenment' 6id Refractory Concretes, EPRI Report

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Chicago, June 1982. a- ;

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80

13. D. A. Campbell-Allen-et al., "An Investigation of the Effect of

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1971.

18. H. L. Malhotra, "The Effect of Temperature on the Compressive

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19. J. C. Marechal, "Variations in the Modulus of Elasticity and Poisson's Ratio with Temperature," SP-34, vols. 1-3, Papei SP 34-30,*

pp. 547-64 in Concrete for uclear Reactors, American Concrete

Institute, Detroit, 1972.

20. K. W. Nasser and R. P. Lohtia, "Mass Concrete Properties at High

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21. S.,Ohgishi et al., "n-Properties of Magnetite and Serpentine Concrete at Elevated Temperatures for Nuclear Reactor Shielding,"

SP-34, vols. 1-3, Paper SP 34-57, pp. 1243-53 in' Concrete for Nuclear Reactors, American Concrete Institute, Detroit, 1972.

22. F. J. Roux, Concrete at Elevated Temperatures, Ph.D. thesis, University of, Capetown, South Africa, August 1974.

23. N. G. Zoldners, "Effect of High Temperatures on Concrete Incorporating Different Aggregates," Am. Soc. Test. Mter. Proc. 60, 1087-1108

(1960).

24. Behavior of Concrete Under-Temperature Extremes, Publication SP-39,

American Concrete Institute, Detroit, 1973.

25. D. J. Naus, Task 2: Concrete-Properties in Nuclear Environment -

A Review of Concrete Material Systems for Application to Prestressed

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Div., Oak Ridge Natl. Lab., May 1981.
81

26. R. Kottas et al., "Strength Characteristicsof Concrete in the Temperature Range'of 200 to 200 0C," Paper H/2,' vol. H Session 1,

Paper 2, 5th Int'Z. Conf. on Structural Mechanics in 'Reactor Technology, Berlin, Germany, Aug. 13-17, 1979.

27. U.S.' Bureau'of-'Reclamation,: A 10-Year Study of Creep Properties of

Concrete,' Concrete Laboratory Report-No. SP-38, Denver, July 28,

1953.

28. Z.' P. Bazant-and S. Prasannan, High-Temperature TrixiaZl Torsional

Creep Tests of Concrete at Various Hygral Conditions,-Report No. 85-

9/285 h, Center for Concrete and Geomaterials, The Technical Institute, Northwestern University, Evanston, Ill., September 1985.

29. Z. P. Bazant et al., Large Triaxial-Torsional Testing'Machine 'ith

Hygrothermal Control, Report No. 85-7/285 L, Center for Concrete and

Geomaterials, The'Technical'Institute, Northwestern University,

Evanston,'Ill., July'1985.

30. G. E. Troxell et al., Long-Time Creep and Shrinkage Tests of Plain

and Reinforced Concrete; ASTM Proc. 48, American Society for Testing

and Materials, Philadelphia, pp. 1101-20,- 1958.

31. D. Campbell-Allen and P. M. Desai, "The -Influence of Aggregate on -

the Behavior of Concrete at Elevated Temperature," NucL. Eng.' Des.

6(1), 65-77 (August 1967).

32. V. V. Bertero and M. Polivka, "Influence of Thermal Exposure on

Mechanical Characteristics 'of Concrete," SP-34, Paper 28, pp. 505-31

in Concrete for'NucZear Reactors, American Concrete Institute,

Detroit, 1972.

33. G. C. Carette et al'., "Sustained High Temperature Effects on Concretes Made 'with Normal Portland Cement,'Normal Portland Cement and

Slag, or Normal Portland Cement and Fly Ash," Concr. Int. 4(7),

41-51 (July 1982).

34. A. P. Meirs; "Long Term-Tests on the Effect'of Moderate Heating on

the Compressive Strength and Dynamic Modulus of Elasticity of Concrete," SP-34', Paper 20; pp. 355-75 in Concrete forNuceear

'Reactors, AmericanConcrete Institute, Detroit, 1972.

35.' C. DeFigh-Price et'al'., Effects-,f Long-Term Exposureto Elevated

Temperature on Mechanical Prbperties of Concrete, ACI Symposium,

Concrete and Cementitious Materials for Radioactive Waste'Management, New York, Nov. 1, 1984.

36. -M. P.-Gillen et !a., Strength nd Elastic' Properties- of Concrete Exposed'to Long-Term Moderate Temperatures and High Radiation Fields,

RHO-RE-SA-55 P, Rockwell Hanford Operations, Richland', Wash.,'1984.
82

37. -H. Kagami et al., "Properties of Concrete Exposed to. Sustained

Elevated Temperatures," 3rd Int'l. Conf. on Structural Mechanics in

Reactor Technology, Paper H/5, London, 1975.

38. A. F. Milovanov and G. D. Salmanov, "The Influence of High Temperature Upon the Properties of Reinforcing Steels and Upon Bond

Strength between Reinforcement and Concrete," Issledovanija po

zharoupornym betonu i zhelezobetonu, pp. 203-23 (1954) (in Russian).

39. C. B. Oland and J. P. Callahan, Bond Between Concrete and Steel Reinforcement at Temperatures to 1490C (3000F), ORNL/TM-6086, Union

Carbide Corp. Nuclear Div., Oak Ridge Natl. Lab., April 1978.

40. U. Diederichs and U. Schneider, "Bond Strength at High Temperature,"

Mag. Concr. Res. 35(115), 75-64 (June 1981).

41. E. G. Peterson, Shielding Properties of Ordinary Concrete as a unction of Temperature, HW-65572, Hanford Atomic Products Operation,

Richland, Wash., Aug. 2, 1960.

42. S. Miyasaka et al., "Gamma-Ray Leakage Through a Slit in Concrete

Shield," Nippon Genshiryaku akkaishi 11, 2-8 (1969) (in Japanese).

43. F. Seboek, Shielding Effectiveness of Cracked Concrete, Kerntechnik

12, 496-501 (November 1970).

44. H. K. Hilsdorf et al., "The Effects of Nuclear Radiation on the

Mechanical Properties of Concrete," SP-55, Paper 55-10, pp. 223-51

in DougZas McHenry, International Symposium on Concrete and Concrete

Structures, American Concrete Institute, Detroit, 1978.

45. C. A. Negin et al., Extended Life Operation of Light-Water Reactors:

Economic and TechnoZogical Review, EPRI NP-2418, vols. 1 and 2,

Electric Power Research Institute, Palo Alto, Calif., June 1982.

46. "Code for Concrete Reactor Vessels and Containments," Sect. IIIDiv. 2, Nuclear Power Plant Components, American Society of MechanicaZ Engineers Boiler and Pressure Vessel Code, New York, 1983.

47. Specification for Prestressed Concrete Pressure Vessels for Nuclear

Reactors, BS5975, British Standards Institution, London, July 1973.

48. V. B. Dubrovskij-et al., The Effect of Neutron Irradiation on Certain Properties of Refractory Concretes," At. Energ. 21, 108-12

(1966). -

49. M. R. Elleuch, F. Dubois, and J. Rappenau, Behavior of Special

Shielding Concretes andTheir Constituents Under Neutron Irradiation," Fourth United Nations International Conference on the Peaceful Uses of Atomic Energy, 1971.
83

50. B. S.' Gray, "The Effect of Reactor Radiation on Cements and Concrete, Commission of the European Communities, Luxembourg;" pp. 17-

39, 1972.

51. T. C. Hansen and L. Eriks'on, "Temper'ature Change Effect'on Behavior

of Cement Paste,'Mortar and Concrete Under Load," Am. Concr. Inst.

Proc. 63(4), 489-504 (April 1966).

52. J. A. Houben, DeBestraZing van MorteZproefstukken [Radiation of

Mortar Specimens], .Commission-'6f therEuropean Communities, Brussels,

1969, pp. 170-83. ' ' Br'ssels,

53. B. T. Kelly and I. Davidson,' "Irradiation Effects on Concrete,"

Group C Paper 16, pp. 173-5,in Conference on Prestressed Concrete

Pressure VesseZs at Church House, Westminster, S.W.I., Institution

of CiviZ Engineers,' London March 13-17, 1967, 1968.'

54. W. Meyer and H. K. Al-Dujaili; "Radiation Effects on Concrete and

Carbonate Compounds," Second Cairo Solid State Conference, Cairo,

1973;.

55. A. ederson, Radiation Damage in Concrete - Measur'ements 'on Mniature Specimen of Cement Mortar,'Commission of the European Communities, Luxembourg, Belgium, 1971, pp. 5-16.

56. B. T. Price, C. C. Horton, 'and K. T. Spinney,- Radiation Shielding,"

pp. 276-78 in InternationaZ Series 'of Monograph on"NucZear' Energy,

Pergamon Press, New York, 1975. -

57. C. R. Van der Schaaf, Effect-of Heating -and Radiation on Some Prop-

'erties of Mortar and Concrete Specimens with Different Compositions,

Commission of the European Communities, Brussels, 1967, pp. 42-71.

58. J.' F. Sommers, "Gamma Radiation-Damage 'of Structural Concrete Immersed in Water,"- Heath Physics, Pergamon Press, Vol. 16, 1969,

pp. 503-8. :':

59. B.' Stoces "et al., The Effect of Radiation o the MechanicaZ Poperties of Concrete,,Ceskoslovenska.Akademie, Translated from the

Czech; Oak Ridge'National''Laboratory, Purchase Order: 34B-83481,

Letter Release No.:- T81, STS'No.: 14087. 

60. R. L. Walker and, M. Grotenhuis, A Sumnary of Shielding Constants for

' Concrete, ANL- 6443, ArgonneNational.Laboratory, Argonne,-'Ill',

November 1961' ' ' ' '

61. M. F.' Eleuch and F. Dubois, "Les Experiences d'Irradiation 'des

Betons et de leurs Constituants," Commission of the European Com-

'' munities, Brussels, 1972,"pp. 716' - ' " muni -* -i " .7 !' -:

62. S. Granata and A. Montaginini, "Behavior of Plastic Mortar Samples

Under Temperature and Radiation Conditions," Information Exchange
84

Meeting Concerning the Results of Concrete Irradiation Programmes,,

18 pp., 1971.

63. B. T. Kelley et al., "The Effects of Reactor Radiation on Concrete,"

-pp. 237-66 inProceedings of the Second Conference on PCPVs and

Their Thermal IsoZation, Commission of the European Communities,

Brussels, November 1969.

64. D. C. McDowall, The Effect of Gamma Irradiation on the Creep Properties of Concrete, Eratom Report No. EUR 4751, pp. 55-69, Commission of European Communities, Luxembourg, 1971.

65. R. G. Clark, Radiation Damage to Concrete, HW-56195, Hanford Atomic

Products Operation, Richland, Wash., Mar. 31, 1958.

66. T. V. Blosser et al., A Study of the Nuclear and Physical Properties

of the ORNL Graphite Reactor hield, ORNL-2195, Union Carbide Corp.

Nuclear Div., Oak Ridge Natl. Lab., August 1958.

67. M. F. Elleuch et al., "Effects of Neutron Radiation on Special Concretes and Their Components," SP-34, vols. 1-3, Paper 34-51,

pp. 1071-1108 in Concrete for Nuclear Pressure Vessels, American

Concrete Institute, Detroit, 1972.

68. H. S. Davis, "N-Reactor Shielding," SP-34, vols. 1-3, Paper 34-52,

-pp. 1109-61 in Concrete for Nuclear Pressure Vessels, American Concrete Institute, Detroit, 1972.

69. S. Granata and A. Montagnini, "Studies of Behavior of Concretes

Under Irradiation," SP-34, vols. 1-3, Paper 34-53, pp. 1163-72 in

Concrete for Nuclear Pressure Vessels, American Concrete Institute,

--Detroit, 1972.

70. -S. C. Alexander,. Effects of Irradiation on Concrete, Final Results,

Atomic Energy Research Establishment, Harwell, England, 1963.

71. A. W. Ch. Batten, Effect of Irradiation on Strength of Concrete,

Atomic Energy Research Establishment, Harwell, England, 1960.

72. G. Christiana et al., Behavior of Plastic Mortar Samples Under Temperature and Radiation Cohditions, Commission of European Communities, Luxembourg, 1972.

73. E. Crispino et al., "Behavior of. Concrete in the Presence of Thermal

Stresses and Radiation," 2nd Information Meeting on Prestressed Concrete Reactor Pressure Vessels and Their Thermal Isolation, Commission of the European Communities, Brussels, November 1969.

74. E. Crispino et al., "Irradiation Effects on Reactor.Structural Materials," Fourth United Nations International Conference on Peaceful

Uses of Atomic Energy, 1971.
85

75. Danish Atomic Energy Commission; Radiation Damage in Concrete - Measurements on Miniature. Specimens of Cement Mortar,-Communications

Research Establishment Riso, Reactor Engineering Department,

Roskilde, Denmark, 1971.

76. R. N. White and P. Gergeley, "Prestressed Concrete Pressure Vessels," short course, Cornell University, Ithaca, N.Y., June 16-20,

1969.

77. P. K. Mehta and B. C. Gerwick, Jr., "Cracking-Corrosion Interaction

in Concrete Exposed to Marine Environments,"-COncr. Int. 4(10),

45-51 (October 1982).

78. D. Darwin et al., "Debate: Crack Width, Cover and Corrosion,"

Concr. Int. 7(5), 20-35 (May 1985).

* . . , , ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I -

79. A. W. Beeby, "Cracking, Cover and-Corrosion of Reinforcement,"

Concr. Int. 5(2), 35-40 (February 1983).

80. T. Z. Harmathy and W. W. Stanzack, Elevated Temperature Tensile and

Creep Properties of Some Structural and Prestressing Steels, ASTM

STP 464, American Society for Testing and Materials, Philadelphia,

1970.

81. A. Cowan and R. W. Nichols, "Effect of Irradiation on Steels Used in

Pressure Vessels," Group D Paper 20, pp. 229-35 in Prestressed Concrete Pressure Vessels, Marilyn S. Udall, ed., The Institute of

Civil Eng., London, 1968.

82. C. P. Tan, Prestressed Concrete in Nuclear Pressure VesseZs, A

Critical Review of Current Literature, ORNL-4277, Union Carbide

Corp. Nuclear Div., Oak Ridge Natl. Lab., May 1968.

83. Technical Report - An International Survey of In-Service Inspection

Experience ith Prestressed Concrete Pressure Vessels and Containments for Nuclear Reactors, FIP/3/5, FIP Commission on Concrete

Pressure and Storage Vessels, Federation Internationale de la Precontrainte, Wexham Springs, Slough, U.K., 1982.

84. D. J. Naus, An Evaluation of the Effectiveness of Selected Corrosion

Inhibitors for Protection of Prestressing Steels in PCPVs, ORNL/TM6479, Union Carbide Corp. Nuclear Div., Oak Ridge Natl. Lab., March

1979.

85. H. Ashar and D. J. Naus, "Overview of the Use of Prestressed Concrete in U.S. Nuclear Power Plants," Paper H1/1, 7th International

Conference on Structural Mechanics in Reactor Technology, Chicago,

August 1983.

86. W. Podolny, Jr. and T. Melville, "Understanding the Relaxation in

Prestressing," PCI J. 14(4) (August 1969).
86

87. J. R. Libby, Modern Prestressed Concrete, Design Principles and Construction Methods, Von Nostrand Reinhold Co., New York, 1971.

88. T. Cahill and G. D. Branch, "Long-Term Relaxation Behavior of Stabilized Prestressing Wires and Strands," Group D, Paper 19, pp. 219-28 - in Prestressed Concrete Pressure VesseZs, Marilyn S. Udall, ed., The

Institute of Civil Eng., London, 1968.

89. S. Freedman, "Properties of Materials for Reinforced Concrete,"

Handbook of Concrete Engineering, ed. M. Fintel, Van Norstrand Reinhold Co., New York, 1974.

90. Product Data Sheet, Visconorust 2090P-4 Casing Filter, Viscosity Oil

Company, Chicago, 1975.

91. R. R. Funk, "Testing Anchorage to Concrete," Civ. Eng. Nuci. Power

1, 6-1-1 to 6-1-17 (Sept. 15-17, 1980).

92. D. F. Meinheit and F. D. Heidbrink, Behavior of Drilled-In Expansion Anchors," Concr. Int. 7(4), 62-66 (April 1985).
87

5. CURRENT TECHNOLOGY'FOR DETECTION OF

- CONCRETE AGING PHENOMENA

Tests are conducted on concrete to assess future performance of a

structure as a result of.(1) noncompliance of strength..tests; (2) inadequacies in-standards for.placing, compacting, or curing of concrete in

the structure; (3) damage resulting from overload, fatigue, frost, abrasion, chemical attack, fire, explosion, or weathering;* and (4)'concern

about the capacity of.the structure to withstand design, actual, or projected loading conditions Item_(4)-of this'-list is of interest'to the

present study because it pertains to life extension:considerations.

Information presented in Chap. 4 indicates that the ability of a

concretecomponent!to continue to meet its functional and performance-requirements over an extended period of time is dependent on the durability

of its constituents. Techniques for.detection of.concrete component degradation should, therefore, address-evaluation of the.concrete, mild

steel reinforcing,- prestressing system, and-anchorage embedments. In the

following sections, the various methods for inspection of concrete materials are described, and recommendations are provided on techniques.-to.be

used in the evaluation of light-water reactor (LWR) concrete components.

5.1. Evaluation of Concrete Materials

Sources of distress that 'are present or can occur in concrete materials include (1) cracking, voids, and delamination and (2) strength

losses. Although not an aging-related phenomenon, whether the concrete

was cast having the specified mix composition could also become-a life

extension consideration.

5.1.1 Detection of concrete cracking, voids;,

* and delamination - .

Discontinuities in concrete structures can-.be.detected by visual inspection, nondestructive testing, or examination-of cores. ,

- 5.1.1.1 Visual-inspection.- Periodicyvisual-examinations of exposed concrete provide a rapid and-effective methodifor-identifying and

defining areas of distress (i.e.,,cracking, spalling, volume change, or

cement/aggregate interaction). By locating, marking, identifying by type

and orientation, and -measuring and recording conditions associated with

the cracks -(seepage, differential movement edge.spalling,:etc.), ahistory that will be of assistance in identifyingrthe cause-and.establishing

whether a crack is active or dormant can be established. A crack comparator capable of width determinations to an accuracy of 0.025 mm can be

used to establishcracks that' are above acritical'size required to permit the entry of hostile environments to attack either the concrete or

*Prolonged exposure to elevated temperature and irradiation conditions should be added to this list for nuclear applications.
88

its steel reinforcement.*2,3 Subsurface cracking, delaminations and

voids, and the extent of cracking, however, cannot be established through

visual examinations.

5.1.1.2 Nondestructive testing. Nondestructive techniques that can

be utilized to determine the presence of internal cracks, voids or delaminations, and the depth of penetration of cracks visible at the surface are available.' These techniques are generally ultrasonic, but

acoustic impact, radiography and radar, and thermal'techniques have also

been-used.

Ultrasonic and-stress wave methods.' Detection of cracks or voids in

concrete by using ultrasonic through-transmission measurements is based

on the principle that the-amplitude and direction of travel of ultrasonic

compressional pulses propagating through concrete will be-changed -when

they encounter acrack.4 The ultrasonic pulses are emitted by a transducer, and the transit time to a receiver is measured by electronic means

in'terms of either transit time (microseconds) or path length. Crack

widths >0.0254 mm are'detectable because transmission across air-filled

voids of this size has been shown to be negligible.5 Large internal

flaws in concrete can be detected by an abnormally long transmission time

and/or a large decrease in amplitude of the ultrasonic pulses as they

pass around a crack. Primary advantages of the technique are that it is

an excellent method for rapidly estimating the quality and uniformity of

concrete and that a low level of user expertise is required to make measurements. Disadvantages are that sound transmissions through concrete

are influenced by a number of conditions (Fig. 43, Ref. 6), and quantitative interpretation of results is difficult.

Sonic coring, a'form f ultrasonic testing, has been proposed as a

method for'detdcting construction faults in concrete pressure vessels and

for detectiiig faults in concrete shields.7 The method consists of

lowering transmitter and 'receiver probes to the bottom of adjacent tubes

(i.e., prestressing tendon conduits or drilled core cavities), filling

the tubes with water for coupling, and slowly raising the probes (20 cm/s)

with the signals continuously'monitored by photographic means, using

modulation of light intensity to represent signal intensity. An area of

bad concrete will be indicated by deflection of the first wave. Although

this method has been successfully used for pile and diaphragm wall construction quality-control,'its application to LWR concrete components is

presumed unlikely Ibecause of the'large number of tendons and'the requirement to fill tubes with water as a couplant and because the scattering

effects of multiple layers of reinforcement would make data interpretation extremely difficult. - -

The pulse-echo technique, which provides an alternative to the

through-transmissic'n-methods,8 is based on monitoring the interaction of

acoustic (or stress)-waves with the internal structure of an object.9 An

*Critical crack widths'for entry of a hostile environment vary significantly de'pending on exposure conditions.''Reference 2 notes that corrosion of steel does not occur in concrete exposed to severe conditions

having crack widths 0.2 mm and in protected conditions where the crack

width is 0.3 mm.'-Additional information on tolerable crack widths as a

function of exposure condition is presented in Ref. 3.
89-

ORNL-DWG 86-4583 ETD

(a) SOUND CONCRETE

SOUND WAVES TRAVEL SHORTEST DISTANCE -:_ _ _ BETWEENTx _ _ _ _ AND _ !!J Rx

- ~~~~Vo- L

. { ) . . - . to ,.

Ib) .REINFORCING STEEL

e: ___ _ _ _ _ _ _ Z,- SOUND WAVES TRAVELLING THROUGH STEEL

ARRIVE BEFORE THOSE THROUGH CONCRETE

lb)

- _ _ - _ ( {c} SHALLOW CRACK

SOUND WAVES TRAVEL AROUND CRACK l * . _ -~~~~tj >to; VI < Vo- .- 

(c) >tV<V-

(dl NARROW CRACK

SOUND WAVES PARTIALLY REFLECTED AND

.. . - + PARTIALLY TRANSMITTED WITH LARGE LOSS

(d) . OF AMPLITUDE BUT ONLY SLIGHT INCREASE

e)WIDE CRACK.

- - . SOUND WAVES WHOLLY REFLECTED, NO

(e _SIGNAL RECEIVED

61 oa o° ae 0 o D * ' 1 (f&g) VOIDS AND MICRO CRACKS

-__________.___-.___-.__-__ .-SOUND WAVES MAY TRAVEL AROUND OR

ACROSS THEM (IF WATER FILLED? WITH

I | 1 fl ! ji 1 | . OF .INCREASE TRANIT TIME

; g) - ; : . - . -

CONDITIONS THAT INFLUENCE TRANSMISSION OF SOUND WAVES IN CONCRETE

Fig. 43. Conditions that influence transmission of sound 'in con-,

crete. SoUrce: V. R. Stirrup et al., "Pulse Velocity as .a Measure of

Concrete Compressive Strength,". InSitu/Nondestructive Testing of Concrete, SP-82, Paper 11, American.Concrete Institute, Detroit, 1984.;'

acoustic pulse is introduced into the test object by either an electromechanical transducer (pulse'repeatability good) or mechanical impact

(simple). Concrete defects are-:detected by-a reduction in the penetrating ability of the high-frequency-waves. The primary advantage of this

technique relative to the through-transmission is that only one face of

the member needs to be accessible because the reflected signal is used.

Principal difficulties intapplication of the technique to concrete are

that the' concrete heterogeneity prevents direct application of the

methods developed for-inspecting metallic structures--a transducer producing both a highly p'enetrating and;relat'ively narrow'ultrasonic beam

his-not been perfected, and interpretation of-results-can be difficult.

'"Acoustic:-emissions:are small-amplitude elastic stress waves gen

erated during material deformation resulting'-from a-mechanical or thermal

stimulus.'-The stress waves are detected by transducers as small-displacements-on the specimen surface.!-Acoustic-emission has been.applied

to concrete for almost 30 years to detect (and locate) distress (cracking) in-concrete components. -Because acoustic'emissions are indicators

of increasing stress levels in,''and--potential subsequent deformation of,

a structure, they can potentially be used to nondestructively determine

the degree of damage that a structure has experienced; that is, the
90

method, therefore, can potentially be used in evaluating the remaining

integrity of a structure that has been subjected to an extreme loading

condition or in estimating the in-service ability of a structure to carry

new loads in-excess of those anticipated during its original design.11

Quantitative application of the technique is difficult, however, because

it requires an understanding of the basic mechanisms that generate microseismic waves within structures and materials; knowledge by which the

disturbances propagate.through a structure; and development of sophisticated instrumentation to accurately identify the nature, severity, and

location of the source.

Acoustic impact methods. Acoustic impact methods, in which the

concrete surface is struck with a hammer, rod, chain, etc., can be used

to detect the presence of'defects through frequenicy and damping characteristics of the "ringing." A portable electronic version developed by

the Texas Highway Department is capable of detecting delaminations up to

66 mm below a concrete surface.12 Advantages of the technique are that

a low level of expertise is required for use and the method does not involve complicated electronic.instrumentation. Disadvantages are that experience is required to interpret results and results are affected by

geometry and mass of the test.object.

Radiography.and penetrating radar methods. The radioactive methods

(X- and gamma-ray techniques) are potentially promising for determining

concrete density, locating reinforcement, and identifying concrete honeycombing. Applications of X-ray radiography in the field, because of its

relatively high initial cost and limited mobility of testing equipment,

have been limited to establishing rebar location, investigating bond

stress in prestressed concrete, and showing''concrete'density variations.

Gamma-ray radiography, because of its use of less costly portable equipment and its ability to make measurements up to concrete thicknesses of

450 mm, has been more widely used to determine position and condition of

reinforcement, voids in grouting of posttensioned prestressed concrete,

voids in concrete, and variable compaction in concrete. 13 Advantages of

gamma-ray radiography are that it uses portable and-relatively inexpensive equipment (relative to X-ray radiography) and can detect internal

defects in a number of materials.. Disadvantages are that the radiation

intensity cannot be-adjusted (thus,-long exposures may be required), it

uses potentially dangerous radiation, and operators must be highly

trained and licensed.'

Penetrating radar using electromagnetic.energy.in the 100- to

1200-MHz frequency range can be used for nondestructive evaluations of

concrete. The waves propagate through the concrete until a boundary

(materials with different dielectric properties) is intercepted; then

part of the incident energy is. reflected, picked up by a receiver, and

indicated by a change in wave shape.. Radar traces are easily able to indicate voids and severely deteriorated material.14 Advantages of the

technique are that large areas of-concrete can be rapidly surveyed and

internal construction details.and.defects identified. Disadvantages are

that where material-differences are small - such as a crack in-sound

material or a contact delamination - transmission differences are hard to

detect and evaluate and material permittivity must be known to determine

the interface depth.:
91.

Thermal methods. Heat-sensing devices are used to detect irregular

temperature distributions caused by the presence of flaws or inhomogeneities in a material or component that has different' impedances to -

heat flow. Contours of equal temperature (thermography) or temperature

levels (thermometry)-are measured over the test surface-with contact or

noncontact detection devices. A common detection device is an infrared

scanning camera. Advantages of the technique are that it is portable, a

permanent record can be made, testing can be done without direct access

to-the surface, and large areas can be rapidly inspected. 'Disadvantages

are that the equipment is costly, reference standards are required, and

moderate to extensive operator expertise is required..

5.1.1.3 Examination of cores. Visual and nondestructive'testing

methods are effectiveinidentifying-areas of- concrete exhibiting distress but often cannot quantify the extent or nature of the distress.

Cores obtained from these areas provide the only direct means to evaluate

the width and depth of-cracking or the extent of voids.

5.1.2 In-situ concrete strength determinations

In conventional civil engineering structures little' attention is

given to the in-situ concrete strength because 28-d (or older) moistcured control specimens are used to indicate the correct strength in a

particular structure,* and very few concrete structures actually-fail.15 1 6

However, for-a structure that'is being considered for extension beyond

its designed service life, especially a structure that has been subjected

to a less than ideal operating environment, the in-situ strength of the

concrete takes on a new meaning. Available methods used to evaluate the

strength -of concrete in a structure include both direct (testing of core

specimens) and indirect techniques (ultrasonic pulse-velocity,-surface

hardness, rebound, penetration, pullout, and breakoff).

5.1.2.1- Direct techniques. Testing of core samples in conformance

with Ref. 17 requirements provides a direct method for obtaining the

in-situ concrete strength. The effects of various factors (core diameter, slenderness ratio, location, etc.) are presented in Refs. 18 and -

19. As noted in Ref.,20,-current American'Concrete Institute statistical

standards relative to the number of:tests required to-ensure that the

probability of -obtaining a-strength-less -than desired is below- a certain

level are not-applicable for in situ tests'. 'Such standards'should' be

developed to achieve reliability of in-situ strength results. -

-5.1.2.2 Indirect techniques.. Indirect techniques measure, some

property of concrete from-which an estimate of the strength is made

- *In-situ strength of 28-d concrete is'niormally significantly less'

(20 to 25%) than 28-d standard control specimen strength of the same concrete because of different compaction and curing conditions.1 Also, systematic. variation of concrete occurs in a structure because of segregation that can reduce concrete strength at the top -of a lift by 15 to 30%

(Ref. 16). - -
92

through correlations that have been developed.*2 1. Nondestructive testing

techniques considered as indirect measures of strength include those

based on surface hardness, penetration resistance, pullout resistance,

break-off resistance, and ultrasonic pulse velocity.

- Surface hardness methods. Three test methods (Williams testing

pistol, Frank spring hammer, and Einbeck pendulum) have been developed

in which the increase in hardness with age of concrete is used to indicate compressive strength.' These methods are all based on the principle

of.impacting the'concrete surface by using a given mass activated by a

given energy and then measuring the size of the indentation. Although

all of these methods are simple to use and provide a large number of

readings in a short time, frequent calibration is required, cement type

may affect results, and strength can generally be determined with an accuracy of only 20 to 30%.

Rebound methods. The Schmidt rebound hammer is basically a form of

surface hardness tester in which a spring-loaded weight is impacted

against the concrete surface and a rebound number is obtained. Concrete

strength is then determined from a manufacturer-supplied chart or from a

laboratory-generated calibration chart. The primary usefulness of the

device is in assessing concrete uniformity in situ, delineating zones (or

areas) or poor quality ordeteriorated concrete in structures, and indicating changes with time of concrete characteristics. Test requirements

are contained in ASTM C 805 Standard Test Method for Rebound Number of

Hardened Concrete. 22 Advantages of the technique are that user expertise

requirements are minimal and' a'large amount of data can be'developed

quickly and inexpensively. Disadvantages are that test results are affected-by concrete surface conditions and the technique only provides a

-rough" indication ''of compressive strength.

Penetration methods.' This type of test, as described in ASTM C 803

Tentative Test Method for Penetration Resistance of Hardened Concrete,23

involves measurement of the resistance of concrete to penetration by a

steel probe driven by-a given amount-of energy. The most common device

of this type is the Windsor'Probe, consisting of a powder-activated driving unit'that propels a hardened alloy probe into the concrete and a

depth gage for measuring penetration. Compressive strength is determined

through calibratiozicurves.-' Advantages and disadvantages of the technique' are essentially the same as for the rebound methods.

-'Pullout resistance methods. Pullout tests, in-conformance with requirements provided in ASTM C-900 Tentative Test Method for PuZZout

Strength of Hardened Concrete.24 involve a determination of the force required to pull'a steel'iinsert out of concrete. Essentially, the method

provides a measure'of the shear strength of concrete, which'is converted

to tensile or compressive strength through correlations. Most of the

pullout methods, such as the LOK-test, require embedment of the metal

pullout insert in fresh concrete.20 Testing of hardened concrete is done

*The primary aplication of these tests is to indicate-differences

in concrete quality from one part of a structure to another, thus indicating areas requiring closer examination through drilling cores and conducting petrographic studies.21
93

using the CAPO test, which is similar to the LOK-test except that'the insert is drilled and expanded wherever required in situ.25 Hardened concrete can also be tested by using'a circular, probe bonded by epoxy resin

to either.a cored or uncored concrete surface26 and by the BRE internal

fracture test27 in which a 6-mm-diam hole is'drilled into the.concrete, a

wedge anchor placed into 'the hole, and the'torque required,'to.pull the

anchor bolt out of,the concrete determined. Advantages of-these methods

are that they are one of the-only nndestructivemethods that'directly

measure in-place'strength and they are economical and rapid. Disadvan-

.tages are that they do not'measure the interior strength of mass concrete

and'theyresult'in the 'requirement for'minor concrete surface repairs.

Breakoff resistance methods. In-situ concrete compressive strength

can be determined from the breakoff strength.of concrete cores formed'

either by plastic inserts while the concrete was'plastic or by drilling

hardened concrete. Bending force applied at right'angles to the'top of

the core at the point of rupture is taken as a measure of the concrete

flexural strength, which is then related through calibration curves to

the compressive strength.28.' Small' cores can then be taken to the laboratory for further examination. Advantages and disadvantages are similar

to those .for the pullout resistance methods.,_,

Ultrasonic pulse velocity methods. Ultrasonic pulse velocity methods

are based on the fact that the velocity of sound in a material is related

to the elastic modulus and material density. Because the pulse velocity

depends only on the elastic properties of the material and not.on the

geometry, itis a very convenient technique for evaluating concrete

quality (i.e., concrete quality proportional to pulse velocity). Procedures and apparatus for determining pulse velocity through concrete are

contained in ASTM C 597 Standard Test Method for PuZse Velocity Through

Concrete.29 Advantages of the technique are that it is a rapid and costeffective method for measuring in-situ concrete uniformity, the method is

totally nondestructive, and it can be utilized to estimate" in-situ concrete strength within 15 to 20% if a good correlation curve has been developed.* Disadvantages are that results are affected by contact surface

smoothness, pulse velocity is somewhat path dependent, pulse velocity can

be affected by temperatures outside the range of 5 to*30°C, the presence

of steel bars parallel to transmission path affects results, and for a

given pulse velocity the compressive strength is.higher for.6ider specimens.30' 31

5.1.3 Mix composition analysis of hardened concrete

Questions concerning whetherthe concrete in a structure.was cast

using thespecified mix.composition can be answered through examination

of core samples. 30 32 .By usinga "point count"'method described in.ASTM

C 457 Standard Recommended Practice for Microscopical Determination of

Air-Void Content,33 the nature of the air void system (volume and.spacing) can be.determined by examining under a microscope a polished section

*Utilization of ultrasonic pulse'velocity to detect'cracks and voids

in concrete was discussed in Sect. 5.1.1.2.
94

of the concrete. An indication of the type and relative amounts of fine

and coarse aggregate, as well as the amount of cementitious matrix, can

be determined by using'ANSI/ASTM C 856 Standard Recommended Practice for

Petrographic Examination of-Hardened Concrete.34 Cement content can be

determined chemically by using ANSI/ASTM C 85 Standard Test Method for

Cement Content of Hardened Portland Cement Concrete.35 Determination of

the original water-cement ratio is not covered by an ASTM standard but

can be estimated by using a British Standard (BS 1881, Part 6)36 that determines the volume of capillary pores originally filled with water and

the"combined water (original water is the sum''of these two). A standard

method also does'not'exist for-determining either the type or amount of

chemical admixtures'used. With respect to mix composition for concretes

that have aged considerably, the determinations are more difficult, especially if the concrete has been subjected to leaching by chemical attack or carbonation.

5.2 Evaluation of Mild Steel Reinforcing Materials

The primary source of distress to which mild teel reinforcement

could be subjected would'be corrosive attack.* Implications of safety

and serviceability of structures undergoing appraisal as a result of rebar corrosion should consider effects on three levels: (1) effect on rebars themselves (cross section or property reductions), (2) development

of fine hairline cracks'in concrete' cover parallel to rebars (indicates

deterioration), and (3) structural cracking or voids (preferential corrosion sites). Safety implications of reinforcement corrosion depend primarily on the structural form or system of construction; second, on the

way in which the~;geometry of the structural components may be affected;

and, third, to alesser extent, on the total amount of 'corrosion of the

rebars.37 Techniques available for corrosion monitoring and inspection

of steel in concrete include (1) visual inspection, (2) mechanical and

ultrasonic tests, (3) core sampling and chemical and physical tests,

'(4)-potential mapping, and (5) rate of corrosion probes.38 ,39

5.2.1 Visual inspection

Visual inspection generally provides the first indication of a corrosion problem. Buildup of corrosion products around reinforcement will

eventually reach appoint where the internal tensile forces generated form

hairline cracks in the concrete following the line of the reinforcement.

Rust staining and concrete spalling also occur as corrosion progresses.

Chipping of the concrete cover to expose the rebar will indicate the

degree of corrosion and may provide clues to its cause.

*Mild steel reinforcement in LWR concrete safety-related components

under normal operating conditions probably would not be subjected to

levels of irradiation or elevated temperatures sufficiently high to produce'a reduction in properties.
95

5.2.2 Mechanical and ultrasonic tests,

Surface tapping using techniques described previously for detection

of delamination can be used to define the area potentially affected by '

corrosion. If a rebound or Schmidt hammer is used to impact the surface,

comparative information may also be obtained on concrete quality.

Ultrasonic pulse velocity measurements also may be used to detect

areas experiencing cracking or delamination caused by corrosion.

5.2.3 Core sampling and chemical and physical tests'-

Cores obtained from'areas indicating distress-as determined by

either of the two previous techniques provide a direct'method for-examining and assessing the extent of'corrosion.' -Apachometer,'or-cover '''-

meter,* can be used to detect the-presence, and-in some cases the depth

or size of reinforcement, so that the corecan be obtained without further damaging the steel. ' i -

Chemical analysis for chloride or sulfate distribution can be conducted on samples-obtained by coring or from dust-obtained by drilling.-

Measurement ofvdiffusion parameters for-oxygen and chloride ions provides

an indication of the ease with which contaminants-enter the concrete.

Areas (depths) that are alkaline and, thus, able to protect the-reinforcement can be identified by using phenolphthalein.-lConcrete composition and-:performance can also be indicated through

electrical resistivity measurements. A high value of resistivity -

(>12000 -cm) indicates that corrosion from galvanic effects is of reduced threat.39

5.2.4- Potential'mapping

Information'on the passivity of reinforcing steel can be obtained

through corrosion potential measurements by using a reference electrode

placed on the concrete surface and connected by means of a high-impedance

voltmeter .(>109 ) to the reinforcement. The-probability of anarea exhibiting corrosion decreases as the-half-cell potential measurement :approaches zero. More details are provided in ANSI/ASTh C.876 Staidard. -

Test Method for Half CeZl PotentiaZs of Reinforcing Steelin Concrete.40

5.2.5 Rate of corrosion probes

'Two types 'of probes that can be embedded into concrete to-provide an

indication of the'rate-of'corrosion are available. 41 The"first- ty'pe 'uses

two or 'three electrically 'isolated 'short sections "of steel i'iie or reinforcing steel and linear 'polarizati6oi tchniques.o 'The'se'cond 'devi'ce'u'ses

'*he presence f steel'affects the magnetic field of a probe with

*Th 'effpresence tthe 

the 'effecti icreasing with proximity 'of thesteel.
96

a steel wire or hollow cylinder embedded into concrete to provide cumulative rate of corrosion data from periodic measurements. The primary application of these devices has been to evaluate the effect of rehabilitation procedures on the corrosion rate.

5.3 Evaluation of Prestressing Steel Materials

The ability of a prestressed concrete containment to withstand the

loadings that would develop as7a result of a loss-of-coolant accident

depends on the continued integrity of the tendons. In the United States

the condition and functional capability of unbonded posttensioning systems must be periodically assessed. This is accomplished, in part, systematically through an in-service inspection program that must be developed-and implemented for each containment. Requirements for containment

tendon surveillance programs in the United States are presented in

(1) Regulatory Guide (RG) 1.35 (proposed revision 3),4 (2) RG 1.35.1

(proposed),43 (3) ASME (proposed Subsection.IWX),44 and (4) U.S. NRC

Standard Technical Specification for Tendon Surveillance.45

The present basis for conducting tendon inspections is presented in

RG 1.35 nservice Inspections of Ungrouted Tendons in Prestressed Concrete Containment Structures (Rev. 2.46 The intent of RG 1.35 is to

provide the utilities with a basis for developing inspection programs and

to provide reasonable assurance, when properly implemented, that the

structural integrity of the equipment was being maintained. Basic components-covered in the RG include sample selection, visual inspection,

prestress monitoring tests, tendon material tests and inspections, and

inspection of filler grease.

Tendon sample selection criteria are specified for typical prestressed concrete containments. If no problems are uncovered during the

first three surveillances (scheduled 1, 3, and 5 years after the initial

structural integrity test), then the criteria for sample selection are

relaxed. In all cases, the tendons are to be selected on arandom but

representative basis.

Anchorage'assembly hardware (stressing washers, shims, wedges, and

bearing;plates) of 'all'tenfidons selected for inspection are to' be examined

visually. (During the integrated leak rate test, while the containment

is'at its maximum test' pressure, visual examination of the exterior of

the concrete is also performed to detect areas of widespread concrete

cracking, spalling, or grease leakage.)

Stress levels of each of the tendons in the sample selected for inspection are monitored by performing lift-off or other equivalent tests.

These tests include measurement of the tendon force level, by using properly: calibrated jacks and the simultaneous measurement of elongations.

Acceptance criteria for the results state that the prestress force measured for each tendon should be within the limits predicted for the time

of the test.

Previously stressed tendon wires or strands from one tendon of each

type are to' be removed from the containment for examination over its entire length to determine if there is evidence 'of corrosion or other deleterious effects. At least three samples are cut from each wire or strand
97,'

(each end and midlength). and tensile tests conducted. At successive inspections, samples should--be.-selected from different tendons.

A sample of grease from each tendon in the surveillance is to be

analyzed (impurities and amounts) and- the results -are to be compared -with

the original grease specification- Also, the presence of voids in the

grease is-to be noted. 

Additional information on in-service inspections of ungrouted tendons in prestressed concrete containment structures-and containment leak

rate testing criteria can be obtained from Refs. 47 and 48, respectively.

5.4 Evaluation of Anchorage Embedments

Failure of anchorage embedments will generally occur as a result.,of

either 'improper'installation or deterioration of the concrete within'-

which it is embedded. Visual inspections can be used to evaluate the

general condition of the concrete near an embedment and to provide a cursory examination of the anchor or anchor plate to check for improper

anchor embedment, weld-or plate tearing, plate rotation, or plate buckling. Mechanical tests can be used -to verify-that pullout and torque

levels of embedments are in conformance with minimums required by design.

Welds or other metallic components can be inspected by using magnetic

particle and liquid penetrant techniques for surface eaminations,"'or if

a volumetric examination -is required, radiographic, ultrasonic, and-eddy

current techniques are available. - -

5.5. Recommended Techniques'for Concrete

Component Inspection

In the previous sections, available-techniques for inspection of

concrete materials were discussed and their advantages and limitations

presented.: Generally, evaluation of concrete components will involve a

combination of several'techniques'(i.e., a visual examination followed by

coring in areas exhibiting distress). Table-6,presents a summary of nondestructive evaluation techniques that are recommended for investigation

of concrete components properties. Once an area of distress is identified, core samples should be obtained to-provide quantitative information

on the extent of degradation, cause, and need for repair. Recommended

and alternate methods for structural integrity monitoring of concrete

comp:nents are also presented in the table. '-.'

., - - C 
98

Table 6. Recommended nondestructive evaluation methods for

inspection of concrete materials

-aterial a.d Available methods of detection Material and.

characteristic Recommended Alternates

Concrete

General quality

Cracking/voids

Strength

Ultrasonic pulse

velocity

Rebound hammer

Penetrating probe

Visual inspection

Ultrasonic pulse

velocity

Acoustic impact

Penetrating probe

Rebound hammer

Pullout methods

Ultrasonic pulse echo

Gamma radiographya

Ultrasonic pulse echo

Gamma radiographya

Breakoff methods

Surface hardness methods

Mild steel reinforcing

Location/size

Corrosion

Pachometer

Gamma radiographya

Visual inspectionb

Electrical potential

measurements

Ultrasonic pulse echo

Penetrating radar

Rate of corrosion probes

Prestressing tendons

Loads

Corrosion

Concrete embedments

Structural integrity

Tendon liftoff tests

Visual inspections

Mechanical property

tests

Tendon load vs

elongation tests

Visual inspections

Mechanical testing

Proof testing

Load cells

Corrosion inhibitor

analysis

Acoustic emission

aLimited to concrete thickness 450 mm.

bReflected through cracking and staining

surface.

observed at concrete
99 

References

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100

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101

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Tester," SP-82, Paper 14, pp.. 277-92 in In Situ/Nondestructive

Testing of Cncrete, American Concrete Institute, Detroit, 1984.

29. "Standard Test Method for Pulse Velocity Through Concrete," ASTM

C 597, Part. 14, Annual Book of ASTM Standards,-American Society for

Testing and Materials, Philadelphia, 1979, pp. 353-56.

30. S. indess and J. F. Young, Concrete, Prentice-Hall, Englewood

Cliffs, N.J., 1981.

31. H. W. Chung, "Effect of Embedded, Steel Bars Upon Ultrasonic Testing

of Concrete," Mag..Concr..Res. 30(102), 19-25 (March 1978).

32. K. Mather, "Preservation Technology: Evaluating Concrete in Structures," Concr. Int. 7(10), 33-41 (October 1985).-

33. "Standard Recommended Practice for Microscopical Determination of

Air-Void Content," ASTM C.457,Part 14, Annual Book of ASI Standards, American.Society for- Testing and Materials, Philadelphia,

1979, pp. 465-69.

34. "Standard Recommended Practice for Petrographic Examination of

Hardened Concrete," ,ANSI/ASTM C 856, Part 14, Annual Book of ASTM

Standards, American Society for Testing and Materials, Philadelphia,

1979, pp. 514-32. .

35. "Standard Test Methodfor Cement Content of Hardened Portland Cement

Concrete," ANSI/ASTM C 85, Part 14, Annual-Book of ASTMStandards,

American Society for Testing and Materials, Philadelphia, 1979,

pp. 41-44. -

36. "Methods of Testing Concrete," BS 1881, Part 6, British Standards

Institution, London, 1983.

37. R.-J. Cuirie, "The Implication of -Reinforcement Corrosion for Safety

and-Serviceability of 'Structures," pp. 11-18 in Proceedingsof a

Meeting on Corrosion of Reinforcement in Concrete Construction, ed.

A. P. Crane, Ellis Harwood Ltd. Publishers, Chichester, England,

1983.
102

38.: J. L. Dawson, "Corrosion' Monitoring of Steel in Concrete," pp. 175-92

in Proceedings of a Meeting on Corrosion of Reinforcement in Concrete Construction, ed. A. P. Crane, Ellis Harwood Ltd. Publishers,

Chichester, England, 1983.

39. P. G. Cavalier and P. R. Vassie, Proc. Institution of Civil Engrs.,

Vol. 70, 461-80, 1981.

40. "Standard Test Method for Half Cell Potentials of Reinforcing Steel

in Concrete," ANSI/AST C 876, Part 14, 1979 Annual Book of ASTM

Standards, American Society for Testing and Materials, Philadelphia,

1979, pp. 539-43.

41. ACI Committee 222, "Corrosion of Metals in Concrete," ACI 222R-85,

pp. 3-32 in Proc. Am. Concrete Inst., V.82, N.1, Detroit (JanuaryFebruary 1985).

42. "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete

Containments," Regulatory Guide 1.35 (proposed revision 3), Office

of Standards Development, U.S. Nuclear Regulatory Commission, Washington, D.C., April 1979.

43. "Determining Prestressing Forces for Inspection of Prestressed Concrete Containments," Regulatory Guide 1.35.1 (proposed), Office of

Standards Development, U.S. Nuclear Regulatory Commission, Washington, D.C., April 1979.

44. "Inservice Inspection of Concrete Pressure Components," Subsection

IWX-2400 Examination of Prestressed Concrete Containment with Unbonded Tendons, Section XI-Working Group, American Society for

Mechanical Engineers, New York, 8/81 (draft), ASME BoiZer and Pressure Vessel Code, Section XI Rules for Inservice Inspection of Nuclear Power Plant Components.

45. J.' F. Fulton, "Tendon Surveillance Requirements - 'Average Tendon

Force," Nucl. Eng. Des. 72(3), 303-308 (October 1982).

46. "Inservice Inspections of Ungrouted Tendons in Prestressed'Concrete

Containment Structures," Regulatory Guide 1.35 (rev. 2), Office of

Standards Development, U.S. Nuclear Regulatory Commission, Washington, D.C., January 1976.

47. J. R. Dougan, Evaluation of Inservice Inspections of Greased Prestressing Tendons, ORNL/TM-8278, Union Carbide Corp. Nuclear Div.,

Oak Ridge Natl. Lab., September 1982.

48. J. R. Dougan,. Evaluation of Containment Leak Rate Testing Criteria,

NUREG/CR-3549 (ORNL/TM-8909), Union Carbide Corp. Nuclear Div., Oak

Ridge Natl. Lab., March 1984..
103

6. REMEDIAL MEASURES FOR REPAIR OR REPLACEMENT OF

DEGRADED CONCRETE COMPONENTS*

Objectives of remedial.work are to'restore the component's structural integrity, to arrest the mechanism'producing-distress, and to ensure, as far as possible, that the'cause of distress will not'recur. :

Basic components of a program to meet these objectives include: diagnosis (damage evaluation), prognosis (can repair be made and is'it economical), scheduling (priority assignments), method selection (depends on

nature of distress, adaptability.of.jproposed.method, environment, and

costs), preparation (function of extent and type of distress),.and'application.l With respect to these components, materials for repair',.preparation of concrete for repair,"and repair techniquest willbe discussed.

Also, several examples of structural component performance before and

after repair will be presented to demonstrate the effectiveness of repair

procedures.''

6.1 Materials for Repair of Concrete

A wide variety of materials is available for the repair and maintenance of'concrete. They range from low-viscosity polymers (epoxies and

polyesters) for sealing fine cracks, to very rapid-setting cements (calcium aluminate.andIregulated set) for repairs in the presence-of water,

-to special concretes (fibrous, latex modified, and polymer)'for overlays,

to portland cment'miortar or concrete. Established'remedial measures

generally involve the use of one or more of-the followingmaterials:

epoxy resins, shotcrete, preplaced aggregate concrete, epoxy ceramic

foam, replacement'mortar or concrete, wedge 'anchors and additional reinforcement, and miscellaneous sealant materials.2' 3

6.1.1 Epoxy resins '

Epoxy resins have a number of'advantageous properties that make them

ideally suited-for use in the repair of concrete. They provide a wide

range- of viscosities and cured physical properties, and they provide excellent bond -strength, even'in the presence of moisture. Limitations-of

epoxies include-thermal expansion and modulus of elasticity properties

*Remedial measures pertain primarily to the, concrete material sys-'

tems. Mild steel' reinforcement repair will be addressed at' appropriate

points, but repair of prestressing systems is not addressed because these

systems are designed to be inspectable and replaceable.

lRetrofitting to increase the load-carrying capability of a component through.strengthening'of existing elements 'or element'addition is

also considered. Although'in the'strict sense this is not'a repair technique, there are situations where retrofitting may be required either in

conjunction with a repair procedure or separately.
104

significantly different from concrete; susceptibility to creep; and elevated temperature exposure, which can significantly reduce the strength

of some formulations. Epoxies used in repair work are generally twocomponent systems with mixing done at the time of usage. Although their

pot'life varies with formulation and temperature,'their effective usage

period normally rns from 5 to 30 min. Problems occurring with the use

of epoxies generally result from improper proportioning, contaminated

substratum, excessive'exotherm, or moisture. American Concrete Institute

(ACI) Standards'relating'to bonding of hardened concrete by using'a

multicomponent epoxy adhesivei include: ACI 503.1-79, Standard Specification for Bonding Hardened Concrete,' Steel, Wood,' Brick, and Other MateriaZs to Hardened Concrete ith a Multi-Component Epoxy;4 ACI 503.2-79,

Standard Specification for Bonding Pastic'Concrete'to Hardened Concrete

with aMAti-Component'Epoxy Adhesivte;5 ACI 503.3-79, Standard Specification for Producing a Skid-Resistant Surface on Concrete by Use of a

Multi-Component Epoxy System; 6 and 503.4-79, Standard Specification for

Repairing Concrete with Epoxy Mortars.7 Additional information also can

be obtained from ACI 503R-80, Use of Epoxy Compounds with Concrete.8

6.1.2 Shotcrete

Shotcreting, or gunniting, is concrete that is'applied pneumatically

by'spraying it from'a nozzle'by means'of compressed air.' Application may

be by means of either a dry-mix or wet-mix process. The dry-mix process

involves premixing the cement and sand* and transferring it-to the work'

site through a hose in a stream of compressed air. The water is injected

and mixed with the material as it exists. In the wet-mix'process all ingredients are thoroughly mixed, material is introduced into the chamber

of the delivery equipment, mix is metered into the delivery hose and conveyed to the nozzle, additional air is injected at'the nozzle to increase

the velocity, and the material is jetted from the nozzle at'high velocity. Because the dry mix can be placed at lower water contents resulting in higher strengths and lower shrinkage, it is most commonly

used. A properly installed dry mix will develop good bond strength and

can obtain compressive strengths to 55 MPa (a strength of 27.6 MPa is

commonly obtained). Advantages of shotcreting are that it is an ideal

method for placement of concrete on vertical or steeply sloped surfaces,

formwork is not required, and shrinkage is virtually eliminated. Disadvantages are that the quality of the material applied is highly dependent on the nozzleman, and about one-fourth to one-half of the material

can rebound on impact. ACI 506-66 Recommended Practice for hotcretinglO

presents additional information on materials, equipment, and applications.

6.1.3 Preplaced aggregate concrete

Preplaced aggregate concrete basically involves packing the forms

.with a well-graded coarse aggregate and injecting structural mortar or

*Concrete for shotcreting is generally made with fairly fine aggregate (<lo mm) and sand, but aggregates up to 20-mm maximum size have been

used. Fibrous concrete can also be applied by shotcreting.9
105

grout into the mass to fill the voids. Because the coarse aggregate

particles are in intimate contact with one another and are' generally;

present in greater quantities, preplaced aggregate concrete exhibits only

about one-half the drying shrinkage of conventional concrete. Also,,high

bond strengths develop-with the existing concrete'because of the selfstressing effect of the grout. Utilization of the method, however, requires skill and experience to ensure complete filling of the voids.

Additional information on preplaced aggregate concrete is contained in

Ref. 11.- - ' -

6.1.4 Epoxy ceramic foams

Epoxy ceramic foams are a two-component formulation that, when

properly mixed, will initiate foam generation within <1 mn and'expand

in volume 7 to 20 times if unrestrained.2 An advantage of the epoxy

ceramic foams relative to-conventional expansive-resins is that even when

completely constrained, the maximum pressure developed is only about '

14 kPa. Strengths obtained are a function of the formulation, application procedures, and expansion permitted and generally will range from

0.7 to 34.5 MPa where expansion -of'<50% is' permitted. In addition to the

low pressure buildup on expansion, advantages of epoxy ceramic foams include extraordinary bond strength to most materials, stability even at

high temperatures, and ability to penetrate crackwidths >0.25 mm. Optimal'injection of the'epoxy ceramic foams,-however, requires a high shear

mixer and heating.

6.1.5 Replacement mortar or concrete -

-Portland cement grout or mortar materials are used for localized

patching.:The material is provided as dry as-possible consistent with

good compaction or pumping. -

Machine-mixed concrete of suitable consistency and proportions is

utilized for areas where concrete replacement* is required.12 To provide

compatibility with'the substratum concrete, it is' best to use materials

and mix proportions as close as possible to those used in the original

construction.

6.1.6 Wedge anchors-and additional reinforcement-

-' Often in'the repair or--rehabilitation of concrete structures,:areas

exist where inadequate shearconnection-between concrete and-steel may be

*Improved performance of replacement mortars obtained'by'replacing'up'to-33%'of the mix water with alatex emulsion''

(polyvinylidefie,"styrene-butadienecopolymer, or polyacrylatecopolymer).

Use of the latex emulsion improves workability; provides increased compressive, flexure, and tensile strengths;'p'r6vides 'excellent bond with

existing concrete; reduces shrinkage cracking and absorption; and provides increased freeze-thaw resistance.12 Cost of the latex emulsion

system is its major limitation.
106

present. Wedge-type anchors or grouted anchors may be utilized for this

purpose. Pullout and shear resistance data for the various wedge anchor

systems is provided by the manufacturer based on static load tests.

Where seismic loadings may. occur, however, the capacity of these systems

needs to be lowered. Results presented in Ref. 13 indicate that the

average failure value of anchors under dynamic conditions-is -15% lower

than the value for static loadings.

Conventional mild steel: and prestressing steel materials can be

utilized in the repair of cracked reinforced concrete construction. The

reinforcement is used to provide load transfer across a crack. Prestressing steel is utilized where a major portion of the member must be

strengthened or where cracks must be closed.

6.1.7. Miscellaneous sealant materials

Information on coating and joint sealant materials is provided in

Refs. 14 and 15.

6.2 Preparation of Concrete for Repair

The effectiveness of a repair to concrete is directly related to the

care that was exercised in-preparing the'substratum.' Deteriorated or defective concrete must be removed to expose sound concrete by chipping,

sawing, drilling, scarifying, planing, or using a water jet. Reinforcing

steel that is corroded or has been mechanically damaged should be removed

and replaced. After removal of defective materials, the exposed concrete

surface should be thoroughly cleaned by flushing with high-pressure water

or vacuum cleaning to remove particles or dust. Where formwork is required,.the forms should be-constructed' and installed in accordance with

ACI 347-68 Recommended Practice for Concrete Formwork. 16

6.3 Techniques for Repair of Concrete

Selection of the technique for repair of a concrete structure depends to a large degree on the size, depth, and area of repair required.

Choice of a repair procedure is also predicated on meeting at least one

of the following objectives: restore or increase strength, restore or

increase stiffness, improve-functional performance, provide watertightness, improve appearance of concrete surface, improve durability, and

prevent access of corrosive materials to reinforcement.3 Types of distress requiring repair that could occur in light-water reactor (LWR)

safety-related concrete components include: - cracking, spaling or delamination,'nonvisible'voids,'and fracturing or shattering. In addition,

situations could occur'in which a component could'require retrofitting

because of either a change in performance requirements or overload.
107

6.3.1 Crack repair methods

Nine potential methods for the repair of cracks are identified:3

(1) epoxy injection, (2) routing and sealing, (3) stitching and additional reinforcement,'(4) drilling and plugging, (5) flexible sealing,

(6) grouting, (7)'dry packing, (8)-polymer impregnation, and (9) autogenous healing.

6.3.1.1 Epoxy injection. The use of pressure-injected lowviscosity epoxy resin can bond cracks as narrow as 0.025 mm. After

cleaning to remove deleterious substances, the'cracks are sealed at the

surface by using thixotropic epoxy, thermosetting wax, or cementitious

materials. Injection of epoxy s performed sequentially through preformed plastic injection ports or through unsealed portions-of the crack.

Either an automated proportioning pump in-head mixing device or batch

mixing followed by injection-from a pressurized vessel procedure is used.

Complete and proper injection of through-cracked members requires sealing

and installation of ports on both sides of the member being injected.

Appearance of epoxy material at all port locations ensures- complete filling of a crack. Epoxy-injection is generally limited to cracks-with a

maximum width of -6 mm.

6.3.1.2 Routing and sealing. Routing and sealing is used primarily

to prevent the entry of hostile environments into dormant cracks. The

procedure consists essentially of enlarging the crack along its exposed

face by using a concrete saw, hand tools, or a pneumatic tool ,and sealing

with a suitable joint sealant. The-technique has application to both

fine pattern cracks and larger isolated defects.

6.3.1.3 Stitching and additional reinforcement. Stitching provides

a method for reestablishing tensile force transfer across a major crack.

Holes are drilled on both sides of the crack and U-shaped metallic units,

spanning the crack are'insertedand grouted.. Where crack watertightness

is required the crack should be sealed prior to'stitching.

An additional techniquefor reestablishing the integrity of cracked

sections is to seal the crack, drill-holes at about a 90° angle to the

crack plane, fill the hole and crack plane with epoxy by injection at low

pressure, and place a rebar into the hole. This'technique bonds the

crack surfaces together and also provides reinforcement. Where additional strengthening or crack closure is required, prestressing strand or

bars can be used to apply a compressive force.'The prestressing force is

generally applied through additional anchors that must be provided.

'6.3.1.4 Drilling'and plugging.'- Drilling and'plugging are utilized

to repair cracks that run'in reasonably straight lines and are accessible

only at-one end,'for example; vertical cracks in retaining walls.' The

technique involves drilling ahole', centered on and'following the crack,'

of sufficient size to intersect'the crackalong its full length''and to '

provide'room for sufficient'material to' structurally take the loads.

After cleaning and sealing, the'hole is filled with a grout material to

form a shear key. -

6.3.1.5 Flexible sealing. 'Active cracks 'can be routed out,-

cleaned, and filled with'a suitable'flexible sealant. k bond breaker is'

provided at the bottom of the slot to allow the sealant to change shape.

This repair technique is applicable to areas that are not subject to

traffic or mechanical abuse.
108

6.3.1.6 Grout injection. Wide cracks in mass concrete structures

that cannot be repaired by epoxy injection can be repaired by injection

of either portland cement grout or a chemical grout. The procedure consists of cleaning the crack surfaces, installing grout nipples, sealing

the crack, flushing the crack to clean, checking the seal, and injecting

the grout under pressure. Portland cement grout mixtures consist of

cement and water or cement plus sand and water (larger cracks). Chemical

grouts consist of solutions of two or more chemicals that combine to form

a gel, a solid precipitate, or a foam.'

---6.3.1.7 Dry packing. The dry pack method has a distinct advantage

because it does not require special equipment. Dry packing is used for

defects that have a high-ratio of depth to area and dormant cracks that

have been slotted. After cleaning, a low water-cement mortar is' placed

into the defect and compacted by'tamping'or rodding. Because the patching'material has a low water-cement ratio, its shrinkage is negligible so

that the patch remains tight.

6.3.1.8 Polymer impregnation. Monomer systems can also be used for

effective repair of cracks. Systems that are used for impregnation contain a catalyst or inhibitor and a monomer or combination of monomers.

Polymerization can be effected by catalytic action' irradiation, or heat,

with heat being he''general method used-in conjunction-with concrete repair. The technique involves drying the concrete surface, flooding it

with monomer, and polymerization in place. Large voids or broken areas

of structures in compression zones (beams) can be repaired by first

filling with fine and coarse aggregate and then flooding the-area with

monomer.

6.3.1.9 Autogenous healing. Autogenous healing is a natural process of crack repair that can-occur in the presence'.of'nonflowing moisture and absence of tensile stress (inactive crack). Healing occurs

through carbonation of the calcium hydroxide in the cement paste by carbon dioxide. The crystals that form interlace and twine, producing a

mechanical bonding effect supplemented by chemical bonding between adjacent crystals and between the crystals and the surfaces of the paste

and aggregate. The effectiveness of this technique decreases with age of

the crack.

6.3.2 Spalling or-delamination repair methods

- Spalling or delaminated areas are satisfactorily repaired by a concrete-overlay provided procedures are taken to provide good bond to the

substratum.* The- process consists of-mechanically removing the damaged

or unsound concrete,-thoroughly cleaning the surface, permitting the surface to dry, applying a thin grout layer or bonding agent, and applying a

high-quality portland-cement-based material or an epoxy-based-mortar or

concrete. Where the area is relatively deep and shear transfer is required between the substratum and repair material, reinforcement dowels

may be grouted into the substratum prior to-placement of the repair material. The use of. mesh or additional reinforcement may also be desirable.

*Techniques for repair of areas exhibiting scaling are similar except the depth of repair is much less.
109

6.3.3 Nonvisible void repair methods

Nonvisible voids such as rock pockets, honeycomb,--or excessive

porosity can be repaired by drilling small diameter holes to intercept

the voids, determining the extent and configuration of the void system by

injection of compressed air or water into the void system, or by visual

inspection 'using a borescope,' and, depending on the magnitude of the

delamination, injecting either epoxy resin, expansive cement grout or

mortar, or epoxy-ceramic'foam. Proper injection of the cement grouts requires prewetting of the substratum with excess water removed prior to

injection.

6.3.4 Fractured or shattered concrete repair methods

Where the concrete has been badly fractured or shattered, the defective material (concrete and possibly rebars) must be removed and replaced.

Either (1) machine-mixed concrete of suitable consistency and proportions

to become integral with the base concrete, (2) shotcrete, or (3) preplaced aggregate materials may be utilized to effect the repair. Type K

shrinkage-compensating cement also is frequently used. Supplemental reinforcement and dowels are used to make the repair self-sustaining and to

anchor it to the sound concrete.

6.3.5 Retrofitting (strengthening) methods

Existing structural components can become inadequate due to either a

change in performance requirements or occurrence of an overload condition

(intense seismic event). Under these conditions retrofitting may be required to reestablish serviceability. As noted in Ref. 2, this can be

accomplished by either strengthening of existing elements, addition of

new force-resisting elements, a combination-of element strengthening and

addition, or use of supplemental. connecting devices.

Strengthening of existing elements can be accomplished by increasing

their shear resistance or cohesiveness by grout or adhesive injection,

encasement, or addition'of section. In cases using encasement or section

addition, careful attention must be paid to providing shear transfer and

bond development. This can be effected-through roughening the host concrete surface and the use of shear transfer devices such as grouted

dowels or wedge anchors. Figures'44'-47present examples of methods utilized for strengthening existing shear walls, columns, beams, and foundations.2 -

New elements or the replacement of existing elements also can provide component structural strengthening. Techniques that can be used include replacement of nonstructural -building'interior walls with walls

designed to impart shear-resistance, addition of new-floor and roof diaphragms,.:or foundation augmentation-.(new elements,-additional piles).

Figure 46 presents an example where .anew collector member was added..

Continuity or fixity of nonstructural elements can be provided by

direct bolting or placement of supplemental steel straps that are bolted

in place. Parapets, towers, overhanging cornices, or support fixtures

also can be braced to restore structural integrity by use of structural
110

ORNL-DWG 86-4584 ETD

SH EAR K EYS 

EPOXY SET DOWEL NEW REBAR NEW EPOXY SET DOWELS REBAR EXTENDED THROUGH HOLE

NEW SHOTCRETE IN EXIST PILASTER-EPOXY IN PLACE

PLAN

DRILL& EPOXY 

NEW DOWELS i>

NEW REBAR

NEW SHOTCRETE ;

MROUGH EXIST FLOOR ;

IE REBAR SET IN EPOXY v

SECTION

DRILL EXIST WALLS SET

/NEW DOWELS IN EPOXY

EPOXY GROUT CRACKS

IN EXIST WALL

/o~

DAILLT

CONTINU

.Fig. 44. Typical shear'wall strengthening. Source: J. Warner,

"Methods for Repairing and Retrofitting (Strengthening) Existing' Build- -

ingsj, Workshop'on'Earthquake-Resistant Reinforced Concrete Building Construction, University of California, Berkeley, July 11-15, 1977.
ill

ORNL-DWG 86-4585 ETD

I, I

j ....., , . .. ... ,... ,,,,,.

.-. * DRILL EXISTWAL

SET NEW "U" TIE'

.L & EPOXY

q -- - - --7-

NEW SHOTCRETE

a . II I.I 0

.,,,. . .? .. I. . I ' :': ,.

e . 7..'..~e .

NEW REINF

NEW SHOTCRETE

6 ~~N EW R EIN F

NEW SHOTCRETE

Fig. 45. Typical methods for column strengthening. Source:.

J. Warner, "Methods for Repairing and Retrofitting (Strengtheninig),Existing Buildings, Workshop-on Earthquake-Resistant Reinforced Concrete

Building Construction, Univers ityof California, Berkeley, July 11-15,

1977.
112

ORNL-DWG 86-4586 ETD

~~~~~~~~BOLTS

NEWREBAR

NEW "U"TIES

EPOXY IN PLACE .

NEW SHOTCRETE

*Fig. 46., Typical methods for strengthening beams and new collector

members' Source: J. Wiarner,,"Methods for Repairing and Retro fitting

(Strengthening) Existing Buildings," Workshop on Earthquake-Resistant

Reinforced Concrete Building Construction, University of California,

Berkeley, July 11-15, 1977.
113

ORNL-DWG 86-4587 ETD

Y SET DOWELS

NEW REBAR

7 ENEW CONCRETE

IEW REINF

EW CONCRETE OR SHOTCRETE

7• NEW REINF DRILL *- - THROUGH EXIST

ECONCRETE &

°On.9 EPOXY IN PLACE

NEW SHOTCRETEC

NEW REINF -

, . I

DRILL EXIST CONCRETE , .

& EPOXY IN PLACE

, .0. C

*1

'., II . .- .. . I. . . . I

. ,e, /// .1 .,,,., .

9 *0o o*O.o

;~ . 0 . 0 

Fig. 47. Typical foundation augmentation. Source: J. Warner,

"Methods for Repairing and Retrofitting (Strengthening) Existing Buildings," Workshop on;Earthquake-Resistant Reinforced Concrete Building Construction, University.'of;California, Berkeley, July 11-15,- 1977.

I* - .. . ,. : * , . -
114

steel members that are either bolted in place, secured by embedment anchor

systems, or embedded in replacement mortar or a polymer-based material.

Figure 48 presents an example of an anchorage for parapets and cornices.2

ORNL-DWG 86-4588 ETD

I, -~~

- I ,,,

-,.....

i,,_.

......-

...... ,

NEW WEDGE ANCHORS

OR EPOXY SET BOLTS

NEW STEEL BRACE

NEW THROUGH BOLTS, WEDGE

ANCHORS OR EPOXY SET BOLTS

EXIST WALL

:.i,,,' 

, .'-'i;,t,. .. , ,,, ,.,,, ,,,,, .,.,,,,..,, ,,,,

, NEW BOLTS

- EXIST, NEW OR

STRENGTHENED WALL

Fig. 48. Typical anchorage for parapets and cornices. Source:

J. Warner, "Methods for-Repairing and Retrofitting (Strengthening) Existing Buildings," Workshop on Earthquake-Resistant Reinforced Concrete

Building Construction, University of California, Berkeley, July 11-15,

1977.
115

6.4 Effectiveness of Repairs to Concrete,

Structural Components

An'indication of the effectiveness of techniques used in the'repair

of concrete 'structural components can be provided by examining the performance of several components before and'after repair. Pertinent examples from the literature'that compare prerepair and postrepair performance include: (1) concrete-rebar'bond, (2) reinforced concrete beams

statically and cyclically loaded, (3) concrete joints under static and

dynamic loading, (4) concrete shear walls under fire exposure, and

(5) earthquake-resistant structural wall.

6.4.1 Concrete-rebar bond

Tests were conducted on pullout'specimens and reinforced concrete

beams (shown-schematically in Fig.'49) to investigate the effectiveness

of epoxy'injection in repairing the bond between steel and concrete in

reinforced'concrete members.1 7 The pullout specimens consisted 'of 150-mm

concrete'cubes-containing a single' Grade'60 deformed bar having a diameter of either 12, 16, or 20 mm.-The specimens were tested by applying a

tensile force to the bar'while the c'oncrete cube was restrained and measuring slip of the bar at the free end. After testing, the perimeter of

each concrete crack ws sealed with a'rapid-setting epoxy adhesive, a

structural adhesive'was pressure injected into the cracks, the specimen

was permitted to cure under ambient conditions for about 7 d, and the

specimen was then retested. To supplement the pullout data, two reinforced concrete beams 200 mm x 300 mm x 2 m long and containing a single

Grade 60 deformed bar of 25-mm diameter were loaded in flexure. To determine steel strains during the test, strain gages were attached at two

ORNL-DWG 86-4589 ETD

75~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~* 0150SQUAR. 14 300 300

.- LVDT DIAL GAGE I'

o ~~~~~~~~~~12 mm STIRRUPS

@100mm c/c

_ _T : _ STEEL DEBONDEPIWITH STRAIN GAGE : . .HELIX . ADHESIVE. ... APE|-1 ............... | ,0:

HELIX . ~ ~ - 25 mm STEEL BAR

STEEL BAR . . .

(A) PULL-OUT SPECIMEN (B) R. C.RBEAM

.Fig..49. Details of pull-out and reinforced concrete beam

test. specimens. Source: H. W.- Chung, -"Epoxy Repair of Bond in Reinforced Concrete Members," Proc. J..American Concrete Institute 78(1),

January-February 1981.-
116

notches located in the beam -300 mm from each end. Also part of the bar

between the notch and end of the beam was debonded to ensure that a bond

failure occurred. After beam failure in shear with substantial rebar

slip, it was repaired using.the same procedure as for the pullout specimens. Although the damaged surfaces of rebar embedment were not. totally

penetrated by'the epoxy', thebond strength of the repaired concrete was

not less than the original concrete, and the repaired concrete could resist the same bond stress with less slip than experienced by the original

concrete.

6.4.2 Reinforced concrete beams

Six reinforced concrete beams (Fig. 50) containing a large rectangular opening were loaded eccentrically.18 The size of the rectangular

opening differed either in length. or depth for each beam. At failure

numerous cracks'developed, and concrete crushing occurred at all four

corners of the opening. The beams were repaired by restoring their shape

(straightness), removing all.loose concrete, replacing the crushed concrete with epoxy mortar, sealing exposed cracks, and pressure-injecting a

mixture of low-viscosity resin/hardener through nipples that had been attached during crack sealing. The repaired beams were then retested in

the same manner as the original beams. Results showed that all repaired

cracks did not reopen on reloading (repaired sections were stronger than

adjacent concrete), crack widths in repaired beams were generally smaller

ORNL-DWG 86-4590 ETD

a ~~~~~~~~~~b 200

P 1000 mm t-A~~~~~~~~-

vrS II E 1 ~~~~~~~~~~~~~~~~~~~~doI2 d

dol2

150 .A 3000 mm - J-.150 mm Section A.A

(a) DIMENSION AND LOADING

6 mm o stirrups 6mm ¢ stirrups 6 mm O stirrups

at 100 mm at spacing St at 100 mm

.1 11 any IRUM ll"Xi W rS1 D|| mmO | I|a2-13 ~2-13 mm 

? I I I I I I .[4=F EI ..... gl I I I I I 213 mm 

Gd 2.13 mm 0 mm 6 mm s sirrups 6 stirrups mm 6 mm stirrups SectionB

at 75 mm as-ai b -s at 75 mm

(b) REINFORCEMENT DETAILS

Fig. 50. Dimensions, loadings, and reinforcement details of reinforced concrete beams containing a large rectangular opening. Source:

M. A. Mansur and K.' C. G. Ong, -"Epoxy-Repaired Beams," Concr. Int'Z.

7(10), American Concrete Institute, Detroit, October 1985.
117

than those originally occurring, repaired beams exhibited reduced stiffness because of the presence of hairline cracks that could not be injected, and the repaired beams were stronger than the original beams.

Beams (shown schematically in Fig. 51) were designed to fail'either

in tension or shear to establish the repair capability and consequences

of synthetic resin injection.19 Unidirectional and reversed (cyclic)

loading sequences were applied tokthe beams. The cyclic loading was applied to establish the effect of'an interposed resin layer within cracked

concrete, resulting from wedging'and possible resin fatigue. Prior to

retesting, the cracks were repaired-by'sealing at the concrete surface

and injecting a synthetic resiii(epoxide). Conclusions-from -the investigation were that badly cracked reinforced concrete beams can be--reinstated to load deflection and ultimate load behavior at least as good as

that for an unfailed beam; the repair technique has limitations if the

crack widths are either too great (repair will not hold) or too small

(<0.1 mm); and for the cyclic tests there were no signs, from the performance of either the resin or the shear-cracked beams, that the repair

worsened the situation by creating new cracks as a result of wedging.

6.4.3 Concrete joints

Shear tests were conducted on concrete pushoff specime ns that were

125 x 200 x 660 mm in length. 20 As shown in Fig. 52, each specimen was

composed of two parts: a precast part of 38 MPa and a cast-in-place part

of 32-MPa concrete. The interface between the parts was a rough surface

produced by exposing the coarse aggregate on the precast part to flowing

water before the concrete had set. Specimens either had no reinforcement

across the joint or two- 5-mm-diam mild steel stirrups were provided

across the joint. The specimens were first loaded axially to produce a

shearing effect along the plane of the joint with slip along the interface monitored. After failure, the specimens were-repaired by clamping

the detached parts tightly together, sealing the perimeter of the joint

with rapid-setting adhesive leaving holes for epoxy injection and air relief, and pressure-injecting epoxy into the crack. After curing 3or 4 d

under ambient conditions, the specimen was reloaded to failure.- Investigation results showed that the shear-resistance -of the repaired joint was

at least equivalent to that of the original joint, deformation capability

of original and repaired joints were equivalent, and shear stresses up to

as high as 5 MPa'could be tolerated by the repaired joint (failure may

occur in the adjacent concrete, however).

Dynamic -shear tests were also conducted using the concrete pushoff

specimen shown in Fig. 52.21 Specimens were fabricated, tested, and repaired using the same procedures as described in the previous paragraph,

except the specimens were loaded-by.axial impact-using-a specially built

jack operated by compressed air rather than loaded statically. Load

cells placed on the top and bottom of the specimen were used to measure

impact loads that produced a rate of stressing of,-12,500 MPa/s. The dynamic tests lead to the following conclusions: dynamic shear strength of

repaired joint was at least. equivalent to that ofithe original joint; the

repaired joint can'absorbithe same-amount;of impulse as the original.

joint; and, provided the repair -is' properly done, the repaired joint is

stronger in shear than the adjacent concrete.- -
118

ORNL-DWG 86-4591 ETD

-EIGHT 6 mm LINKS

AT 180 mm CENTRES

I

50 FOUR 10 mm BARS

ni A- . ,' , s Ho . # I - orIC]L I i - 1 I 1 1

12 mm /

DIA.

1200 JT T

PAD- F L R 

PADS FOR REVERSED LOAD TESTS ONLY

I I I HU ~

140 iI

!,- 1400 , .

LOADING PADS FOUR 25 mm DIA. BARS

\TWO 32 mm DIA. BARS

ELEVEN 6 mm DIA. LINKS

AT 125 mm SPACING

FIVE 8 mm DIA. LINKS

AT 250 mm SPACING

12 mm

DIA.

4 Add _ fin~~~~~~~~I 

. zCI n n n IRI n ~~~~~I n n- .I 

I I @*

I F 4 -- .- - .- 1 r=====J -0 -

M - -a J I I . IILl , ii

- lA n '

'Fig. 51. Schematics of:tension and shear crack-type reinforced concrete beiam test specimens.' Source: P. C Hewlett and J. G. D. Morgan,

"Static and Cyclic Response Reinforced Concrete Beams Repaired by Resin

Injection," Mg. Concr. Res. -34(118) (May 1982).
119

ORNL-DWG 86-4592 ETD

LOAD

200 8" )

5 m DIA. TIRR UPS

.. w r - ' (SERIES R ONLY)

10

:o - Z A~~~* 

... . ° -OF- .-- .;.-.......~~~~~~~~~~~~~~~~~~~~~............ .. to

5 mm DIA. REBARS

10 mm DIA. REBARS

Fig. 52. Details of concrete pushoff test specimens. Source:

H. W. Chung and L. M. Lui, "Epoxy-Repaired Concrete Joints," Proc. J. Am.

Concr . Inst. 74(6) (June 1977).

6.4.4 Concrete walls under fire exposure

Results presented in the previous paragraphs indicate that when a

structural component is repaired properly with an epoxy-based system, it

will exhibit equivalent or superior performance characteristics relative

to the original structure. Extremely low-probability environments (lossof-coolant accident), however, could occur in an'LWR plant in which a

concrete component repaired by epoxy injection may be required to meet

its functional and performance requirements under less than ideal conditions, for example, elevated temperature. Because epoxies, like most

materials, exhibit strength reductions on elevated temperature exposure,

the performance of epoxy-repaired structural components under these conditions needs to be established. Some insight' into this problem can be

found in Ref. 22, which presents results of the effects of elevated temperature exposure on "basic" structural epoxy systems and the behavior of

epoxy-repaired concrete shear :walls during "pseudo-fire! exposures.

Pure epoxy adhesive-specimens'12.7-mm diam 'by 25.4-m long were

placed into a preheated electric oven for a period of 1 hat the specified temperature and then tested in compression immediately upon removal
120

from the oven (hot strengths). Companion tests were also conducted in

which the specimens were permitted to cool at room temperature for about

7 d before testing (residual strengths). Above 204'C the epoxy hot

strength was found to be negligible due to cracking and the rubberlike

specimen behavior. Residual strengths up to 149°C exposure were reduced

<25%, but beyond 204'C the specimens cracked and became rubberlike, exhibiting a strength reduction (40% strength reduction at 204C).

Small- (356- by 457-mm), intermediate- (864- by 1016-mm), and largescale (2286- by 2591-mm) prismatic specimens of varying wall thickness

(152.4 to 254 mm) and crack widths (1.27 to 6.35 mm) were fabricated

using 28.6-MPa ready-mix concrete (Fig. 53). Crack surfaces were simulated by breaking each wall specimen as a beam. After curing for 90 d

under standard laboratory conditions, the specimens were epoxy injected

to reestablish integrity. Six structural epoxies, representing materials

that had been used to repair structures-damaged by the San Fernando earthquake, were used in the investigation. The epoxy-repaired shear wall

specimens were then subjected to pseudo-fire exposures designed to simulate a 2-h duration ASTM E-119 fire exposure and, a short-duration highintensity (SDHI) fire. During fire exposure(face ABCD in Fig. 53), the

small-scale specimens were not subjected' to external: loadings, but upon

completion of fire exposure, hot (within 10 min) and residual strength

compression tests were conducted. The intermediate- and large-scale test

specimens were nominally loaded to 1.51 and 0.8 Pa, respectively, during

and after fire exposure. For the 2-h ASTM E-119 and the 1-h SDHI fires,

the properties of epoxy-repaired concrete walls .152 to 254 m in thickness were reduced to levels below original design stress levels. Residual strength properties of most of the epoxy materials subjected to

ORNL-DWG 86-4593 ETO

APPLIED COMPRESSION LOAD

35 mm 3 * 864 mm *: CRACK

2286 mm * IDTH 

F. Geea L. .. . -- C 457 mm D

1516 mm ' ~~~~ ' ~~259 mn

Fig. 53. General configuration of small-scale expoxy-repaired

shear wall specimens. Source: . J. M. Plecnik et al., "Epoxy-Repaired

Concrete Walls Under Fire Exposure," ASCE J. Str. Div. 108(STP8) (August

i982)..
121

elevated temperatures were increased more than 50% as a result of postcuring. The duration and intensity of fire exposure were found to have

great significance on strength and behavior of epoxy-repaired concrete

walls both during and after fire exposure; that is, compressive strength

properties after SDHI fire exposure were about two times greater than-for

the ASTM E-119 fire. Also, the orientation of the epoxy-repaired crack'

in relation to applied stress was found to have a significant effect on

the strength properties of epoxy-repaired components during fire exposure

with cracks subjected to parallel shear stresses exhibiting lowest

strengths.

6.4.5 Earthquake-resistant structural wall

One-third-scale''specimens,- representing five-story walls, were'loaded

laterally through the top slab until web damage' occurred (Fig. 54).23 of

ORNL-DWG 86-4594 ETD

7 95/-0" (22m)

(2.36m) 

-be//- 21 >~~~~~(203mm)

X z *: i P -TOP SLAB

1 i0 l02mmi)s -- .

(4.57m)

-- (I.;1 M d BASE BLOCK

10o0" 

(3.05m)

(122m

Fig. 54. rNominal'dimensions-of''6ne-third-scale five-story wall test

specimens. Source: AE. Fi6rato-et,'al.,"Behavior ofiEarthquake Resistant Structural Walls Before and After Repair," J. American Concrete Institute 80(5) (September-October 1983).
122

the three specimens 'tested, 'only one was loaded axially during testing.

A different repair procedure was'investigatd for each wall: damaged web

was.replaced to its' original thickness with new concrete, web thickness

was increased as part of the repair, and supplementary reinforcement was

added to the web (diagonal bars) prior to replacement'of web concrete to

its original thickness. The specimens were then retested and results

compared with initial wall performance. Conclusions'from the results

presented were that replacement of damaged concrete in webs of'structural

walls is an effective and simple repair procedure that'yields strength

and deformation capacities equivalent to the original walls; initial

stiffnesses of repaired walls were -50% those of original.walls (important for dynamic loadings); for the specimen repaired with a thickened

web, deformation capacity of the wall was increased, nominal shear

stresses at equivalent loads were reduced, and the capacity of diagonal

compression struts that form under lateral load reversals were increased;

and addition of diagonal reinforcement within the hinging region of the

base of the repaired wall reduced shear distortions and increased deformation capacity.

References

1. J. J. Waddell, "Basic Steps of a Concrete Repair Program," Concr.

Int. 2(9), 30-33 (September 1980).

2. J. Warner, "Methods for Repairing and Retrofitting (Strengthening)

Existing Buildings," Workshop on Earthquake-Resistant Reinforced

Concrete Building Construction, University of California, Berkeley,

July 11-15, 1977.

3. ACI Committee 224, "Causes, Evaluation and Repair of Cracks in Concrete'Structures," Report No. 224.R-84, pp. 211-30 in Proc. J. Am.

Conc. Inst. 81(3), Detroit, May-June 1984.

4. ACI Committee 503, "Standard Specification for Bonding Hardened Concrete, Steel, Wood, Brick, and Other Materials to Hardened Concrete

with a Multi-Component Epoxy," ACI 503.1-79, American Concrete Institute, Detroit, pp. 437-41 in Proc. J. of American Concrete Institute, Vol. 75, No. 9, September 1978.

5. ACI Committee 503, "Standard Specification for Bonding Plastic Concrete to Hardened Concrete with a Multi-Component Epoxy Adhesive,"

ACI 503.2-79, American Concrete Institute, Detroit, pp. 442-47 in

Proc. J. of American Concrete Institute, Vol.,75, No. 9, September

1978.

6. ACI Committee 503, "Standard Specification for Producing a SkidResistant Surface on Concrete by Use of a Multi-Component Epoxy System," ACI 503.3-79, American Concrete Institute, Detroit, pp. 448-53

in Proc. J. of American Concrete Institute, Vol. 75, No. 9,

September 1978.
123

7. ACI Committee 503, "Standard Specification for Repairing Concrete

with Epoxy Mortars,, ACI 503.4-79,-American Concrete Institute,

Detroit, pp. 454-59 in Proc. J. of American Concrete Institute, Vol.

75, No. 9, September 1978.

8. ACI Committee 503, Use of Epoxy Compounds with Concrete," ACI503R-80, American Concrete Institute, Detroit, Part 5, Masonry, Precast Concrete, Special;Processes, ACI Manual of Concrete Practice,

1985. ;

9. S. Mindess and J. F. Young, Concrete, Prentice-Hall, Englewood

Cliffs, N.J.,-1981.

10. ACI Committee 506, Recommended Practice for Shotcreting," ACI

506-66, American Concrete Institute, Detroit, Part 3, Products and

Processes, ACI Manual of Concrete Practice, 1977.

11. ACI Committee 304, "Preplaced Aggregate Concrete for Structural and

Mass Concrete," pp. 785-97 in Proc. J. of American Concrete Institute, Vol. 66, No. 10, Detroit, October 1969.

12. P. C. Robery, "Structural Repairs," Concrete 17(2) (February 1983).

13. Static and Dynamic Loadings of 5/8 Inch Concrete Anchors, Report

7745.10-72, Department of Engineering Research, Pacific Gas and

Electric Co., San Francisco, Aug. 10, 1972.

14. ACI Committee 515, Guide for the Protection of Concrete Against

Chemical Attack by Means of Chemical Coatings and Other CorrosionResistant Materials, Part 3, Products and Processes, ACI Manua of

Concrete Practice, American Concrete Institute, Detroit, 1977.

15. ACI Committee 504, "Guide to Joint Sealants for Concrete Structures," Part 3, Products and Processes, ACI Manual of Concrete Practice, American Concrete Institute, Detroit, 1977.

16. ACI Committee 347, "Recommended Practice for Concrete Formwork," ACI

347-68, American Concrete Institute, Detroit, 1968, Part 1 Materials

and Properties of Concrete, Construction Practices and Inspection,

Pavements and Slabs, ACI Manual of Concrete Practice, American

Concrete Institute, Detroit, 1976.

17. H. W. Chung, "Epoxy Repair of Bond in Reinforced Concrete Members,"

pp. 79-82 in Proc. J. Am. Concr. Inst. 78(1) (January-February

1981).

18. M. A. Mansur and K. C. G. Ong, "Epoxy-Repaired Beams," Concr. nt.

7(10), 46-50 (October 1985).

19. P. C. Hewlett and J. G. D. Morgan, "Static and Cyclic Response of

Reinforced Concrete Beams Repaired by Resin Injection," Mag. Concr.

Res. 34(118), 5-17 (March 1982).
124

20. H. W. Chung and L. H. Lui, '!Epoxy-Repaired Concrete Joints," Proc.

J. Am. Concr. Inst. 74(6), 264-67 (June 1977).

21. H. W. Chung and L. M. Lui, "Epoxy-Repaired Concrete Joints Under Dynamic Loads," Proc. J. Am. Concr. Inst. 75(7), 313-16 (July 1978).

22. J. M. Plecnik et al., "Epoxy-Repaired Concrete Walls Under Fire Exposure," ASCE J. Str. Div. 108(STP8), 1894-1908 (August 1982).

23. A. E. Fiorato et al., "Behavior of Earthquake Resistant Structural

Walls Before and After Repair," J. Am. Concr. Inst. 80(5), 403-13

(September-October 1983).
125 -

7. CONSIDERATIONS FOR DEVELOPMENT OFA DAMAGE METHODOLOGY

-TO ASSESS DURABILITY FACTOR DETERIORATION RATES

AND TO PREDICT STRUCTURAL RELIABILITY

Information reviously-presented indicates that -the performance of

concrete components in both-nuclear-and non-nuclear applications has been

very good. Where the concrete in these components has been fabricated

with close attention-to the factors-related to -production of -good concrete (Fig. 21)', the 'concrete will'exhibit infinite durability; however,

where there has been a breakdown in one of the'se'factors'or-the component

was subjected to an extreme-'environmental stressor, distress can occur.

Review of the various nondestru'ctive'and destructive techniques:for identifying and indicating the magnitude of distress in concrete has shown '

these techniques to be capable of locating regions subjected to deteriorating influences. -Also, remedial measures that can lead to'successfulrepair and replacement of-concrete have been shown to be available, provided a procedure- such as that'shown schematically in Fig.' 55 -is-fol- '

lowed. Where the system break's down, however, is: that-a damage methodology to provide a quantitative measure of the ability of a structure -to

meet potentialfuture performance'requirements-[e.g., loss-of-coolant

accident (LOCA)] does not presently exist. Three areas, however, that

would provide significant'inputt toward quantifying the ability of a

light-water reactor (LWR)-safety-related concrete 'component to meet its,

functional and performance'requirements''at some' future time, based on-its

performance history or 'present status, can be addressed: (1) development

of a representative material property 'data base,'(2) -establishment and '

evaluation of an accelerated 'aging 'methodology for concrete materials'

-DRNL-DWG 86-4595 ETD

SELECT PROPER UTILIZE - PREPARE . ... REPAIR - PROPER . . : DAMAGED

. TECHNIQUE M ATERIALS -- AREA

: s - .I. | . , /0

DETERMINE - I .* FOLLOW

CAUSE AND; - - ' ; ' RECOMMENDED

SEVERITY . P E - K- APPLICATION

\s~~WILL PROVIDE A o. 

SATISFACTORY REPAIR

. -n -'. - - M ,f- r, n .........

MEETING REQUIREMENTS FOR

STRENGTH | DURABILITY APPEARANCE ECONOMY

Fig. 55. Factors influencing the successful repair-of a concrete

component. Source:- Modificatio'n of J. J. 'Waddell, -"Basic' Steps of a

Concrete Repair Program," Concr. Int'l. 2(9),'American Concrete Institute, Detroit, September 1980. - '
126

and (3) formulation of. a methodology to provide a quantitative measure of

structural reliability and of residual life.

7.1 Development of Representative Material

Property Data Base

Overall performance of a structure is largely dependent on (1) the

quality. of the materials used, which, in turn, is affected by the standard of workmanship, and (2)-for concrete, a structure's.function, position, and the environmental-stressors to which it may be subjected.

Under normal operating conditions a high.level of confidence can be

placed -in traditional material performance based on past experience.

However, for concrete material systems used in LWR applications where operating- conditions are not.necessarily considered normal because of potential elevated temperature and irradiation exposure over a protracted

time, the confidence.level will not be as high as for-the more traditional applications. This is.not-the result of obvious deteriorating influences operating on these structures, but rather from the lack of a

historical material property data base that can be used to form the basis

for life extension considerations. . - .

Three plants that are currently.shut down (Dresden 1, Humbolt Bay,

and Shippingport),*.however, provide an opportunity- for making major contributions to the material property data base relative-to aging effects.

Baseline information on the concrete materials and control specimen

strength results should be available as part of the quality assurance

(QA) documentation. By obtaining concrete..core samples at pertinent locations in one or more of these plants and conducting petrographic examinations and load-to-failure tests on these samples, an indication of the

significance-of aging can be obtained.

Prestressing tendon in-service surveillance reports. and containment

integrated leak-rate test reports also provide a valuable data source.

Results obtained from scrutinizing these reports would provide significant information useful in trending performance of not only the concrete

materials, but also prestressing materials, corrosion inhibitors, seals

and gaskets, etc. Sufficient data should be available to provide information on deterioration parameters affecting these materials and to provide at least a first cut at establishing durability factors for use in

life extension considerations.

7.2 Accelerated Aging Methodology for Concrete Materials

Prediction of the service life of a building component or material

is dependent on there being either sufficient available data on performance of the component or-material under representative conditions for

*Dresden 1 =200 MW(e) boiling-water reactor (BWR).

Humbolt Bay - 63 MW(e) BWR. -

Shippingport = 72 MW(e) pressurized-water reactor-light-water

breeder reactor (PWR-LWBR).
127

the time period of 'interest or accelerated testing methods that can be

used with confidence to develop the required data within a reasonable

time. Although, as noted in Sect. 7.1, valuable data on aging effects

can be obtained from plants that-have been shut down, the data in all

likelihood will be somewhat plant specific and probably -will not be representative for either allsafety-related concrete components or potential environmental stressors. A possible alternative approach'that can

be used to develop the'required data base is to use accelerated aging

test techniques. Either deterministic or probabilistic analyses can then

be applied to the data to predict service life.

Accelerated aging tests-have been used for many materials, such as

insulation, paints, glasses, polymers, etc., to'predict useful remaining

life or to aid in predicting service life. To a limiting degree, tests

of this type have also been applied to predict, at an early age, the 28-d

strength of concrete (accelerated strength testing),l to predict potential concrete strength 'at any age, 2 to predict long-terim s'ervice life of

concretes in a sulfate environment,3 -and to evaluate resistance of 'concrete to freezing and thawing.4 The analytical-experimental program proposed would be based mostly on the American Society for Testing and Materials (ASTM) practice for developing accelerated aging tests to aid in

the prediction of building component service life (shown schematically

in Fig. 56, Ref. 5). The program would involve three major phases:,'

(1) problem definition (characterization-of material or component, identification of pertinent degradation factors and their method(s) of simulation, and definition of test performance requirements), (2) design and

performance of predictive service life tests (experimental-studies in

which pertinent degradation factors would be simulated at an accelerated

rate and predictive service life tests would be' compared to long-term

tests under service conditions), 'and (3) mathematical model development

(compare rates of change in predictive service life tests with those from

in-service tests; Table 7 presents several mathematical models used in

aging studies).. Table 8 presents the major steps and systematic diagrams

similar to those that would be used for analysis of data obtained from a

prototype accelerated life test.13

Results obtained from this study will aid in describing and understanding the phenomena of potential deterioration with the passage of

time, assist in determining the residual service life 'of materials and

components in conjunction with actual'degradation condition, and'help in

establishing maintenance or remedial measure programs that will assist in

either prolonging a component's service life or improving the probability

of the components surviving an-extreme event, such as a LOCA.

7.3 Methodology to -Provide a Quantitative Measure

of Structural Reliability

Assessment of the functional and performance characteristics of concrete components' is an important consideiation in the extension of the

operating life'of'nuclear facilities. 'Given the complex nture of the

various environmental stressors that can exert deterioratiig influences

on the concrete components, a systems approach is'probably best-in addressing the evaluation of a structure for life extension considerations.
128

ORNL-DWG 86-4596 ETD

PART 1 - PROBLEM DEFINITION

t aietiaof in-use performance cane meh

las degradtion indictorsinduced to aweleatedrin testsial

enesan peroraereqiemnsao 3~~~~~~~~~~ tresictieervice dntfyoctical romne cac4etfthcdtyendrge5Inif o degradation te ott and |hrc'rus

PART 2'- PRE-TESTING

I Design and perform preliminary

accelerated aging tests to demonstrate rapid failures caused by

individually applied extreme degra,

dation factors and to confirm degradation mechanisms

g Design and perform predicive ser- 10 Design and perform long-term

vice life tests using the degradation tests under service conditions

obtafinedb berotn-seriend fo

predictive service I fe tests

PTT PcEhg n o r Are te eT

<idn anu db preditor srveliion

te repare tatie tose oerdmon

QUETION: c are th; hne

in-service?

Yes

PART 4- INTERPRETATION AND

REPORTING OF DATA

13 Develop mathematica modes o

degradation and compare' rates of

vice lifetestsusingthchange in predictive service life tests

with those from in-service tests

I &~~~~~

14 Establish performance criteria r

predictive service life tests

16 Predict service life under expecte

in-service conditions

16 65ortth a adata 

,Fig. 56. ASTH E 632 recommended practice for developing predictive

sevce lie es. Source: "Standard Practice for Developing Accelerated Tests to, Aid,Prediction of the Service Life of .Building Components

and Materials," ASTMHE 632,Part 4, Concrete and ineral Aggregates,

Annua1 Book of Standards, American Society for-Tests and Materials,

Philadelphia, 1979.
dF

Table 7. Several mathematical models used in accelerated aging studies

Model Description

b exp 1 + b5 Model-for weathering of plastic

materials6 -

P.: Ob' + l b x + bX2 X2 + bXX 0 b 1 111 22 2 121X2 Study of temperature and irradiation

effects on a composite7

,_ .

I

I.o

4

i-1

4

biXi + 

J=1

4

F2 bij XiX i 1jj 

P b + b (t'- 250)

Study of irradiation, temperature,

water, and exposure time effects on

a polyethylene and PVC 8

Study of weatherometer testing of

polystyrene, PVC, and cross-linked

polyester

Study'of heat aging of polyacrylonitrite and polychloroprene

elastomer system1 ..

Study of propellent.-lifell

%- * 0

Pt k;

0o 'k 

k = A exp (-B/RT)

.p p + k log t .0

I

I I. . O Study of potting compounds, adhesives,

spiralloy, and pressure seals12 ;

- f = A + B log M Accelerated strength testing of

concrete2
130

Table 8. Major steps and schematic diagrams for analysis of data

obtained from a prototype accelerated life testa

1. Measure degradation rate for each quality x(t) associated with observed

changes in material properties and performance characteristics for generalized stress levels So, S ... Sn where each successive stress level

is of higher magnitude than previous stress level.

PROPERTY

XIt)

* . -

.

0

a0.

: (S < Si ... < S51

TIME It) TIME (t

2. Relate degradation rates to measures of environmental and/or operational

stresses (i.e., Arrhenius type plot).

DEGRADATION RATE

OF THE MATERIAL

OR PERFORMANCE

PROPERTY

(tI

PROPERTY 1

ROPERTY k

1/STRESS NORMAL

STRESS

3. Predict amount of degradation of each quality expected to be observed at

next measurement time.

PROPERTY

x t)

WITHIN ASS

STRESS /

LEVEL /

PREDICTED

AMOUNT

PREDICTED TIME

TIME (t)

PROPERTY

xi(t

BETWEEN SS

LEVEL STES /| PREDICTED

I I12

Ti M t

TIME Ct)

4. Verify predicted degradation amounts using data obtained from subsequent

measurement times

aAdapted from: G. B. Gaines et al., inal Report on Methodology for

Accelerated Aging Tests for Predicting Life of Photovoltaic Arrays, ERDA/JPL

954328-77/1, Battelle Columbus Laboratories, Feb. 1, 1977.
131

Basic components of such an approach would encompass the development of

(l).a classification scheme for structures, elements, and deterioration

causes and effects; (2) a methodology for conducting a quantitative assessment of the presence of active deteriorating influences; and (3) the

structural reliability techniques 'to' estimate the ability of a structure

or component to meet potential future requirements, such as a LOCA.

7.3.1 Component classification scheme

Considerable work toward development of a classification scheme has

been done by the School of Civil and Mining Engineering at the University

of Sydney.14 In the study, flow charts are used to categorize types of

structure, elements, and causes and effects'of durability problems.

Structures are initially categorized according to use (e.g., thermal

.power station) and then broken down by structural classification (e.g.,

building), structural ele'men'itclasificationY(e.g., wall), element subclassification (e.g., shear wall), durability factors (e.g., metallic

corrosion), and types of deterioration phenomena (e.g., rebar general

corrosion). Figure 57 presents an'example of the detail that can be

realized with the University of Sydney categorization system. Measurement, intensity, and distribution factors for each of the deterioration

effects were also developed. Results of this study should have direct

application to LWR nuclear-safety-related concrete components, particularly if the consequences of component failure are also factored into the

study.

7.3.2 Methodology for conducting a quantitative assessment

of-the presence of active deteriorating influences

and-their effects -

Detection of age-related degradation,'as well as its magnitude and

rate of occurrence, is a key factor in maintaining the readiness of

-safety-related concrete components to continue their functions in the unlikely event that.a condition, such as a LOCA, would occur. In-service

'inspection (ISI) requirements are imposed'on nuclear plants through documents such as the following: 10 CFR 50; Nuclear-Regulatory Commission

(NRC) Regulatory Guides; Plant Technical Specifications; Inspection and

Enforcement,.(I&E) Bulletins; NRC letters; and-American Society of Mechanical Engineers (ASME) Boiler and Pressure esseZ Code. 1 5. However, because

each nuclear plant has a different construction permit;docket date, construction permit issue-date,-and operating license issue date, each plant

could potentially have a different set of minimum ISI requirements.

Therefore, to simplify life extension of nuclear-safety-related concrete

components, having a standardized ISI program that could not only be'used

to identify but also to quantify any deteriorating influences would be

advantageous.

Limited information on criteria, inspection, and testing requirements for development of such a procedure are available in the form of

'documents published by the American Concrete Institute: for example,'

Guide for Making a Condition Survey of Concrete in Service,1 6 Strength..

Evaluation of Existing Concrete in Service, 1 7 Practices for Evaluation
USE CATEGORIZATION I

ORNL-DWG 86-4598 ETO

I.

I.

II

a

DI

II

II

I

I

STRUCTURAL

CLASSIFICATION DURABILITY: CAUSE PROCESS DETERIORATION EFFECTS

Il - I -

Fig. 57. Systematic evaluation method for determination of structural durability factors.

Source: Based on material presented in H. Roper, D. Baweja, and G. Kirby, "Towards a Quantitative Measure of Durability of Concrete Structural Members," Paper.-SP 82-32, In Situ/Nondestructive Testing of

Concrete, Publication SP-82, American Concrete Institute, Detroit, October 1984.
133

of Concrete i Existing Massive Structures for Service Conditions,18 and

Guide for Concrete Inspection. 9 Additional information is also contained in Refs. 20-22. The application of requirements presented in

these documents to nuclear-safety-related structures being considered

for life extension, however, needs to be evaluated.

A methodology similar to that 'presented in Fig. 58* needs to be

developed, and criteria need to be established for application to nuclearsafety-related concrete components. Quantification of durability 'factors, such as those presented in Ref. 14, needs to be addressed and input

into the methodology for evaluating-the structural condition of concrete

components'. Limits need to be placed on magnitudes of the deterioration

factors (e.g., crack sizes), and probabilistic techniques should be applied to account for random variations and-uncertainties in the measured

parameters.-that can affect loadings and material strengths. Once this

procedure has been-developed, the issue of determining the reliability

of the structure to meet potential future requirements and/or prediction

of component.service life can be addressed.

7.3.3 Structural reliability technique development for

life extension evaluations

Once it has been established that a component has been subjected to

environmental stressors that have'resulted in deteriorating influencest

the effects of these influences must be related to a structural reliability assessment, especially if the component is being considered for

an extended service life. A methodology for conducting such an assessment presently does not exist. One approach, however, might be to calculate the reliability of the particular component-by using a safety index

factor in conjunction with a damage probability matrix that would characterize the'probabilistic nature of the damage that had occurred or is

expected to occur over the component's anticipated service life..

Shinozuka and Tan have used the damage probability matrix approach

to estimate the reliability of a seismically damaged concrete structure

when subjected to a future earthquake.23 Damage states are defined in

the study,.and conditional, as well as initial, damage probability

matrices are introduced in such.a manner that the definition of damage

is consistent with the kind of accuracy achieved when the extent of

structural:damage is estimated through field inspections. The initial

damage.-probability matrix indicates the probabilities that an undamaged

structure will experience.-various degrees of damage, representative of

the corresponding states of damage, after it is subjected to an earthquake of specified intensity. -The conditional damage probability matrix

used is essentially a Mark6vian transitionmatrix that describesithe

transition probabilitywith which a structure in a certain state !of damage will reach another'state-'after being subjected to an earthquake of a

*Adaptation.of'a procedure presented in Ref. 22.

tor for that.matter, even a structure that exhibited no signs of deteriorating influences would have to be evaluated for life extension considerations. -
ORNL-DWG 86-4597 ETD

STRUCTURAL ASSESSMENT AND REPAIR I LIFE EXTENSION

I CONSIDERATIONS

I~~~~~~~~~~~~~~~~~~~~~~~~~~

I

I

I

I

Il 

*,I

NU \ JdUI ~ * Iu1I Ur .MINAI I N

I ASSESSMENT 

DEVELOP N

ADDITIONAL .

AND * l YES I

TEST DATAl 

I . PROVIDE

_ l _ REQUIRED I

XL I . REPARATIONS I

Fig. 58. LWR concrete component evaluation methodology. Source:

Adaptation of a procedure presented in T. L. Rewarts, "Safety Requirements and the Evaluation of Existing Buildings," Concr. Int'l. 7(4),

-American Concrete Institute, Detroit, April 1985.

, ,
135

given intensity. The validity of using probability damage matrices is

demonstrated analytically by considering a vertically standing, symmetrically'reinforced.concrete cantilever beam subjected to ahorizontal

ground acceleration (modeled as a nonstationary random process). Crack

widths that develop near the beam-fixed end-are'relatedrto damage'ratios

(stiffness change) that develop .under different earthquake intensities.

Further verification is provided by a Monte Carlo simulation with the aid

of a nonlinear dynamic structural analysis involving artificially generated earthquakes.

Results obtained in the above investigation are sufficiently encouraging that the use of a damage probability matrix approach* should be

considered as a method for addressing future structural reliability determinations. Various deteriorating effects (e.g., cracking) resulting

from environmental stressors could be modeled and their effect'on structural performance determined under simulated LOCA conditions.

*Concepts of fracture mechanics as applied to cracked reinforced

concrete structures also may merit consideration.

-References

1. V. Ramakrishman and J. DietzAccelerated Strength Testing,,SSP-56,

Paper 7, American Concrete Institute, Detroit, 1978, pp. 95-116.

2. H. S. Lew and T. W. Reichard, Prediction of Strength of Concrete

from Maturity, S-56, Paper 14, American Concrete Institute,

Detroit, 1978, pp. 229-48.

3., G. L.-Kalousek et al., "Concrete for-Long-Time Service in Sulfate

Environment," Cem. Concr. Res. 2(1) (1972).

4. "Standard Test -Method for Resistance of Concrete to Rapid Freezing

and Thawing," ANSI/ASTM C666, Part 14, Concrete and Mineral Aggregates, Annual Book of Standards, American Society for Testing and

Materials, Philadelphia, 1979.

5. "Standard Practice for Developing Accelerated Tests to Aid Prediction of the Service Life of Building Components and Materials," ASTM

E 632, Sect . 4, 1985 AnnuaZ- Book of, ASTh Standards, American Society

for Testing and Materials, Philadelphia, 1985.-

6. National Bureau of Standards and Manufacturing Chemists Association,

Outdoor Performance 'of Plastics I through X, 1968-1972.

7. M. E. Leikina andG.-0. Talevos'yan, "Planning an Experiment for

Studying the Light Aging of Polyethylene Film, SoV. Plast. (7), 73

(1970).- ,-

8. M. E. Leikina et al., "Use of Mathematical Experiment Planning

Method to-Investigate the Influence of Artificial Weather. Factors on

Certain Properties of Polymer Films," Sov. Past.(2), 41 (1971).
136

9. M. R. Kamal, "Effect of Variables in Artificial Weathering on the

Degradation of Selected Plastics," Polym. Eng. Sci. 6, 333 (1966).

10. D. K. Thomas and R. Sinnott, "Prediction of Shelf Storage Life from

Accelerated Heat Aging Tests," J. IRI, p. 163 (1969).

11. G. E. Myers, Chemical Structural Aging Tests, MANCP Report No. 297,

1974.

12. E. M. Dalaba, Minuteman Stage III Component Report, AFRPL-TR-74-58,

1974.

13. G. B. Gaines et-al., FinaZReport on Methodology for Designing AcceZerated Aging Tests for Predicting Life of Photovoltaic Arrays,

ERDA/JPL 954328-77/1, Battelle Columbus Laboratories, Columbus,

Ohio, Feb. 1, 1977.

14. H. Roper et al., Durability Studies of Concrete Structures Czassification and Quantification, Research Report No. R465, School of

Civil and Mining Engineering, University of Sydney, Australia,

January 1984.

15. D. D. Davis, "Optimizing an Inservice Inspection Program Plan," Nuclear Plant Safety, 37-40 (May-June 1985).

16. ACI Committee 201, "Guide for Making a Condition Survey of Concrete

In-Service," pp. 905-18 in Proc. J. Am. Concr. Inst. 65(11) (November 1968).

17. :ACI Committee 437, Strength Evaluation of Existing Concrete Buildings," J. Am. Concr. Inst. 64(11), 705-10 (November 1967).

18.. ACI Committee 207, "Practices for Evaluation of Concrete in Existing

Massive Structures for Service Conditions," Concr. Int. 1(3), 47-61

(March 1979).

19. ACI Committee 311, "Guide for Concrete Inspection," Concr. Int.

2(10), 81-85 (October 1980).

20. B. Bresler, Essential Steps in Adaptation of old Buildings, Workshop

on Earthquake-Resistant Reinforced Concrete Building Construction,

University of California, Berkeley, July 11-15 1977.

21. D. Kaminetzky, Verification of Structural Adequacy, Workshop on

Earthquake-Resistant Reinforced Concrete Building Construction,

University of California, Berkeley, July 11-15,-1977.

22. T. L. Rewarts, "Safety Requirements and the Evaluation of Existing

Concrete Buildings," Concr. Int. 7(4), 50-55 (April 1985).

23. M. Shinozuka and R. Y. Tan, "Seismic Reliability of Damaged Concrete

Beams, J. Struct. Div. 109(7) 1617-34 (July 1983).
137

8. SUMMARY, CONCLUSIONS, AND RECOMMENDATIONS

8.1 Summary

The objectives of the study were to (1) expand upon the work that

was initiated in the first two Electric Power Research Institute studies

relative to longevity and life extension considerations of safety-related

concrete components in light-water reactor (LWR) facilities and.(2) provide background that will logically lead .to' subsequent development of a

methodology for assessing and predicting'the effects of'aging on the performance of concrete-based materials-and components. These objectives

are consistent with Nuclear Plant Aging Research (NPAR) Program.goals

to: (1) identify and characterize aging'and sei'vice wear effects that,

if unchecked,-could cause degradation'of structures, components, and systems and, thereby, impair plant safety; (2) identify methods of inspection,.surveillance, and monitoring, or of evaluating residual life'of

structures, components, and systems', that' will ensure' timely detection of

significant aging effects before loss, of safety function, and (3) evaluate the effectiveness of storage, maintenance, repair, and replacement'

practices in mitigating the rate and extent of degradation caused-by

aging and service wear. .,. .

.. Applications of.safety-related concrete components to LWR technology

were identified, andpertinent components (containment buildings,'containment base mats,.biological shield.walls and.buildings,.and auxiliary

buildings), as well as the materials of which 'they are constructed (concrete, mild steel reinforcement,,prestressing systems, 'embedments, and

anchorages), were described. Historical performance of'concrete components was established through information 'presented onconcrete'longevity, component behavior in both LWR and'high-temperature gas-cooled

reactor applications, and 'a review.'of probienms'with'concrete components

in both general civil engineering and nuclear power applications. The.

majority of the problems identified in conjunction with nuclear ,power'applications were minor and involved either concrete cracking, concrete

voids, or low concrete strengths t early ages.. Five'incidences involving LWR concrete containments.considered significant were'described in'

detail from occurrence and detection through remedial measures used to

restore structural integrity or continuity. These incidences were

related to design, construction, or-human errorfand involved two dome'

delaminations, voids under tendon-bearing plate6; anchor head failures,

and a breakdown in quality control and construction management.

Potential environmental tressors and aging'factors to'which'LWR

safety-related components could'be subjected'were identified and discussed in terms of durability factors related to the materials used to

fabricate the components (e.g., concrete, mild-steel reinforcement, prestressing systems, and embedments). -The current technology for detection

of concrete aging phenomena was also presented in terms of methods applicable to the particular material system'that could-experience deteriorating effects'. Remedialr'measures'for"the repair'r'or replacement'of degraded

concrete components were discussed,-and examples of prerepair and postrepair stiuctural''performance-wer&'presented to indicate the-effectiveness of these measures. Finally, considerations relative to development
138

of a damage methodology for assessment of durability factor deterioration

rates and prediction of structural reliability were discussed.

8.2 Conclusions

Based on the results of this'investigation, the following conclusions can be derived: .'

1. The performance of concrete-based components in both general

civil engineering and nuclear power applications has been exemplary.

Distress that has occurred was generally due to' construction or material

errors.

2.: Techniques for detecting effects of environmental stressors on

concrete materials are sufficiently developed to provide qualitative

data. However, quantitative interpretation can be complicated because of

either-(a) the requirement'for development of correlation curves; (b) embedment (rebars, anchorages, etc.) effects on measured quantities, such

as time of ultrasonic wave' transmission;, or (c) accessibility. Also, a

methodology for-application of this technology to provide required data

for either structural reliability or'life extension assessments needs development.

3. Remedial measures for repair of degraded concrete components are

capable"of.completely restoring structural integrity wen'proper' techniques'and materials.are used. However, results obtained from shear wallcomponents subjected to fire exposure-after being repaired by structural

epoxies indicate that sme' additional work on development of more temperature-resistant epoxies may be merited.

4. The durability of concrete constructions is' affirmed by the

presence of many structures tat have been in existence for periods of

time ranging from several decades to several millenia; however, welldocumented data on concrete longevity that can be used as a basis for

life extension considerations is almost nonexistent.

5.. Primary effects that could lead'to a' loss of serviceability of

concrete components 'in LWR'plants include concrete cracking and loss ofstrength resultingfrom environmental stressors; however, seveiity criteria (e.g., statistically-based crackfwidth tolerances and corrosion inhibitor impurity levels) for degradation of these'components need to be

established.'

6. A damage methodology to provide a quantitative measure of the

durability of a structure with respect to meeting potential future requirements [e.g., loss-of-coolant accident (LOCA)] does not presently

exist.

8.3 Recommendations

-The following recommendations are made:

'1. Existing facilities that have been shut down after an.extended

period of service (e.g., Shippingport, Dresden 1, and Humbolt.Bay) should

be used to obtain aging-related data for concrete materials. Also, these
139

facilities can be used to evaluate the applicability of various techniques for detecting the effects of environmental stressors (primarily

elevated temperature and irradiation) on the concrete materials. By comparing results from the nondestructive examination/tests with those obtained from core tests, considerable insight can be gained toward evaluating. the ability of these tests to provide quantitative data useful for

residual life assessments.

2. Accelerated aging techniques should be investigated as a method

for supplementing the extremely limited data base on concrete aging.

This technique would also have application to other materials used in

conjunction with concrete.

3. Available prestressing tendon in-service inspection records and

data obtained during containment integrated leak-rate tests should be examined as potential sources of information for trending concrete component behavior. Also, for plants that are likely candidates for life extension considerations (e.g., plants with lengthy construction periods),

consideration should be given to increased emphasis on in-service inspections to provide trending information that could potentially shorten the

process required for life extension evaluations.

4. Criteria on durability factor* significance need to be established.

5. A methodology needs to be developed to provide a quantitative

measure of structural reliability either now or later. Such a methodology would use a systems approach and encompass component classification, techniques for quantitative determination of presence and magnitude

of deteriorating influences, and structural reliability assessments. By

using trending of environmental stressor data (concrete aging), the

scheme would enable (a) an assessment of the ability (probability) of

various safety-related concrete components to meet their design requirements (e.g., LOCA) later and (b) prediction of a component's residual

life. For example, an estimation can be made of the time when the influence of an environmental stressor would produce a decrease in concrete

strength to a value below that specified in the design as necessary to

ensure that the structural component meets normal operating and accident

condition requirements.

Reference

1. B. M. Morris and J. P. Vora, Nuclear Plant Aging Research (NPAR) Program Pan, NUREG-1144, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington, D.C., July 1985.

*Identification of the various deterioration phenomena acting on a

particular structure and the assignment of a weighting factor to each of

the phenomena on its significance relative to life extension considerations.
141

Appendix A

LICENSED U.S. POWER REACTORS AS OF APRIL 30, 1985

[From NucZear Safety 26(4), July-August 1985]
-

Reactora Design power

Reactor Docket-No. type Operating Containment typeb

(designer) MW(t) MW(e) license

Arkannas 1

Arkansas 

2

Beaver Valley 

1

Big Rock Point

Browns Ferry 

I

Browns erry 

2

Browns'Ferry 3

Brunswick'1 '

Brunswick 

2

Byron 10

Callaway 1

Calvert Cliffs '1

Calvert Cliffs 

2

Catawba 1'

Cook .1

Cook 

2

Cooper,,

Crystal River 3

Davis-Besse I .

Diablo Ca yon

Dresden 1

Dresden 2

Dresden 3

Duane Arnold

Farley 1:

Farley:2

Fermi 2 

-

Fitzpatrick

Fort Calhoun

Fort St. -Vrain

Ginna

Grand Gulf '10

Haddam Neck --

Hatch 1

Hatch 

2

Humboldt Bayd

IndianPoint 

2

Indian Point 3

50-313

50-368.

50-334

50-155

50-259

50-260

50-296

50-325

50-324

50-454

50-483'

50-317

50-318

50-413

50-315

50-316

50-298

50-302

50-346:

50-275.

50-10

50-237

50-

249

50-331

50-348

50-364

50-341

50-333

50-285

50-267'

50-244'

50-416

50-213

50-321

50-366

50-133

50-247

50-286

P(BW)

P(CE)

P(West)

B(GE)

B(GE)

B(CGE)

B(CE)

B(CE)

B(GE)'

P(West)

P(West)

P(CE)

P(CE)

P(West)

P(West)

P(West)

B(GE)

P(B&W)

P(B6W)

P(West)

B(GE)

B(GE)

B(GE)

B(GE)

P(West)

P(West)

B(GE)

B(GE)

P(CE)

HTGR(GAT)

P(West)

B(GE)'

P(West)

B(GE)

B(GE)

B(GE)

P(West)

P(West)

2568

2815

2652

240

3293

3293

3293

2436

2436

3425

3411

2560,

2560,

3411

3250:

3391

2831

2560

2772

3338

700

2527

2527

1593

2652

2652

3292

2436

1420

842

1520

3833

1825

2436

2436

220

2758

2760

850

912

852

75

1065

1065

1065

821

821

1120'

1120

845

845

1145

1054

1060

778

802

906

1084

200.

794.

794.

538

829

829

1093

821

457

330'

490

1250

575

786

795

63

873

873

1974

1978

1976

1964

1973

1974

1976

1976

1974

1984

1984

1974

1976

1984

1974

1977

1974

1976

1977

1984

1960

1969

1971

*1974

1977

1980

1985

1974

1973

1973

1969

1982

1974

1974

1978

1969

1971

1975

PC - shllow dome, 3 buttresses

PC- shallow dome, 3 buttresses

RC- subatmospheric

S - spherical

S - MKI

S - MKI

S - MKI'

RC - MKI

RC-MKI

PC-shallow dome, 3 buttresses

PC- hemispherical dome, 3 buttresses

PC - shallow dome, 6 buttresses

PC.- shallow dome, 6 buttresses

S-ice condenser.

RC - ice condenser

RC - ice, condenser.

S -MKI .. .

PC 

- shallow dome, 6 buttresses

S - cylindrical

RC - hemispherical dome

S - spherical

S - MKI

S-MKI

S - MKI.

PC- shallow dome, 3 buttresses

PC - shallow dome, 3 buttresses

S - MKI

S- KI

PC- partial prestress, 0 buttresses

PC

PC - hemispherical dome, 0 buttresses

RC -S MKIII

RC - hemispherical dome

S - MKI

S - MKI

S - cylindrical

RC - hemispherical dome

RC - hemispherical dome

-

"S.
Reactora Design power

Reactor Docket No. type license Containment typeb

(designer) HW(t) MW(e)

Kewaunee

La Crosse

LaiSalle 1

La Salle 2

Limerick le

Maine Yankee

McGuire 1

McGuire 2

Millstone Point 1

Millstone Point 2

Monticello

Nine Mile Point 1

North Anna 1

North Anna 2

Oconee 1.

Oconee 2

Oconee 3

Oyster Creek

Palisades

Palo Verde e

Peach Bottom 2

Peach Bottom 3

Pilgrim 1

Point Beach I

Point Beach 2

Prairie Island 1

Prairie Island 2

Quad Cities I

Quad-Cities 2

Rancho.Seco

Robinson 2

Salem 1

Salem 2

San Onofre1

San Onofre 2

San Onofre 3

Sequoyah I

50-305

50-409

50-373

50-374

50-352

50-309

50-369

50-370

50-245

50-336

50-263

50-220

50-338

50-339

50-269

50-270

50-287

50-219

50-255

50-528

50-277

50-278

50-293

50-266

50-301

50-282

50-306

50-254

50-265

50-312

50-261

50-272

50-311

50-206

50-361

50-362

50-327

P(West)

B(A-C)

B(GE)

B(GE)

B(GE)

P(CE)

P(West)

P(West)

B(GE)

P(CE)

B(GE)

B(GE)

P(West)

P(West)

P(B&W)

P(B&W)

PCB&W)

B(GE)

P(CE)

P(CE)

B(GE)

B(GE)

B(GE)

P(West)

P(West)

P(West)

P(West)

B(GE)

B(GE)

PCO6W)

P(West)

P(West)

P(West)

P(West)

P(CE)

P(CE)

P(West)

1650

165

3323

3323

3293

2560

3411

3411

2011

2560

1670

1850

2775

2775

2568

2568

2568

1930

2200

3817

3293

3293

1998

1518

1518

1650

1650

2511

2511

2772

2200

3423

3423

1347

3410

3410

3423

535

50

1078

1078

1065

790

1180

1180

660

830

545

610

898

907

887

887

887

650

805

1250

1065

1065

655

497

497

530

530

-789

789

918

700

1090

1115

430

1100

1100

1148

1973

1973

1982

1984

1984

1972

1981

1983

1970

1975

1971

1974

1977

1980

1973

1973

1974

1969

1971

1984

1973

1974

1972

1970

1971

1973

1974

1971

1972

1974

1970

1976

1980

1967

* 1982

1982

1980

S - cylindrical

S - cylindrical

PC - MKII

PC - MII

RC - MKI1

RC - subatmospheric

S - ice condenser

S - ice condenser

S -.MKI

PC - shallow dome, 3 buttresses

S - KI

S - MKI

RC - subatmospheric

RC - subatmospheric

PC - shallow dome, 6 buttresses

PC - shallow dome, 6 buttresses

PC - shallow dome, 6 buttresses

S - KI

PC - shallow dome, 6 buttresses

PC - hemispherical dome, 3 buttresses

S - MKI

S - MKI

S - MKI

PC - shallow dome, 6 buttresses

PC - shallow dome, 6 buttresses

S - cylindrical

S - cylindrical

S - HKI

S - MKI

PC - shallow dome, 3 buttresses

PC --RC hemispherical dome, 0 buttresses

RC - hemispherical dome

RC - hemispherical dome

S - spherical

PC - hemispherical dome, 3 buttresses

PC - hemispherical dome, 3 buttresses

S - ice condenser

W
Reactora Design power Operating C

Reactor Docket No. type license Containment typeb

(designer) MW(t) MW(e)

Sequoyah 2 50-328 P(West) 3423 1148 1981 S - ice condenser

St. Lucie 1 50-335 P(CE) 2560 802 1976 S - cylindrical

St. Lucie 2 50-389 P(CE) 2560 810 1983 S - cylindrical

Summer I 50-395 P(West) 2775 900 1982 PC - shallow dome, 4 buttresses

Surry 1 50-280 P(West) 2441 822 1972 RC - subatmospheric

Surry 2 50-281 P(West) 2441 822 1973 RC - subatmospheric

Susquehanna 1 50-387 B(GE) 3293 1050 1982 RC - MKII

Susquehanna 2 50-388 B(GE) 3293 1050 1984 RC - MKII

Three Mile Island 1 50-289 P(B&W) 2535 819 1974 PC - shallow dome, 6 buttresses

Three Mile Island 2 d 50-320 P(B&W) 2772 906 1978 PC - shallow dome, 6 buttresses

Trojan 50-344 P(West) 3411 1130 1975 PC - hemispherical dome, 3 buttresses

Turkey Point 3 50-250 P(West) 2200 693 1972 PC - shallow dome, 6 buttresses

Turkey Point 4 50-251 P(West) 2200 693 1973 PC - shallow dome, 6 buttresses

Vermont Yankee 50-271 B(CE) 1593 514 1972 S - MKI

Washington NP 2 50-397 B(GE) 3323 1100 1984 S - MKII

Waterford 3e 50-382 P(CE) 3410 1113 1984 S - cylindrical

Wolf Creek 1 50-482 P(West) 3411 1150 1985 PC - hemispherical dome, 3 buttresses

Yankee Rowe 50-29 P(West) 600 175 1961 S - spherical

Zion 1 50-295 P(West) 3250 1040 1973 PC - shallow dome, 6 buttresses

Zion 2 50-304 P(West) 3250 1040 1973 PC - shallow dome, 6 buttresses

ap pressurized-water reactor

B - boiling-water reactor

B&W - Babcock and Wilcox

CE - Combustion Engineering

West - Westinghouse

GE - General Electric

CAT - CA Technologies Inc.

A-C - Allis Chalmers

bpc prestressed concrete

RC - reinforced concrete

S - steel

C1n power ascension phase.

doperating license but shut down indefinitely.

eLicensed for low-power testing.

.P- Un
147

Appendix B

ANNOTATED LISTING OF PROBLEM AREAS ASSOCIATED WITH

CONCRETE COMPONENTS IN LWR APPLICATIONS
II

Year of Year of

Reactor plant Docket No. commercial of 'Summary description ' '

operation occurrence I I

Yankee Rove

San Onofre I

Ctnna

Indian Point 2

Turkey Point 3

Turkey Point 4

Palisades

Fort St. Vran

Oconee 213

Peach Bottom 2/3

50-29

50-206

50-244

50-247

50-250

1961

1968

1970

1974

1972

1967

1976

1981

1974

1968

1970a

.1974

1975

1982

A 46-u (15-ft) shrinkage crack, covered with fiberglass andrecoated

Voids at 14-locations in diesel generator building center wall; areas from 0.09 m2

(I ft2) with 7- to 10-cm (3- to 4-tn.) penetration to several square meters

(square feet) with full penetration; repaired with dry pack, grout, or concrete

Excessive loss of prestressing, tendons retensloned with no recurrence noted in

subsequent inspections

Concrete temperature local to hot penetration >66'C (150F) but <930C (200'F), no

safety problem due to relatively low periods of. exp sure

Voids below containment wall and near reactor pit, repaired with high-strength

grout, guniting, or dry packing

Dome delamination; delaminated concrete removed, additional rebars provided,

concrete replaced - -

Crease leakage from 110 of 832 tendons at casing,' tendon casings repaired and'

refilled

Concrete spalling at horizontal joint at containment ringgirder with cavities

3 to 5 cm (I to 2 in.) wide by 7 to 10 cm (3 to 4 In.) deep, no threat to

structural integrity, repaired by dry packing

Small void under equipment hatch barrel-, no threat to structural Integrity,

repaired by grouting - ' '

50-251 1973

50-255

50-267

197 1

1979

50-270/287 1974

50-277/278 1974

1981 Approximately 0.1 3 (0.4 ft2) of concrete with nadequate'fines,'area removed and

refilled with concrete 

1975 Sixty-three out of 3780 bttonheads inspected found split, no threat to structural

Integrity

1984b Tendon wire failures noted because of tendon corrosion caused-by micreklological

attack of corrosion inhibitor, analysis revealed sufficient tendons intact to

provide structural integrity, and- surveillance ncreased and tendons nerted by

nitrogen blanket

1982 nuring final ractor building interior -nspectton, two vertical tendons in secondary shield wall of unit 2 were found failed and some tendons in units 2 and 3

were exhibiting corrosion near 'stressing washers; tendons are not required to meet

shield wall functions but were replaced and bottom grease caps redesigned to permit water drainage; surveillance was increased

1983 Four tendons in reactor building found ungreased, tendons inspected and grease

applied - . .

1969 Alumtnum pipe used to place concrete caused concrete strength reduction up to 502,

low-strength concrete in biological shield wall and floor slab of turbine

building replaced - I . .- - I

1979 - Cracking in-concrete supports for two heat'exchingers-caused by thermal expansion

of heat exchanger shells, cracks repaired and supports modified

. Surry I -

. u. 

50-280 1972
year of Year of

Reactor plant Docket No. commercal of Summary description

operation occurrence

Three Mile Island 1 50-289 1974 1975 Two of six concrete footings for rigid pipe supports cracked due to design defi1974

ciency, footings were replaced using a new design

Cracking <0.02 cm (<0.010 in.) wide In containment building ring girder and around

tendon bearing plates, cracks repaired and monitored during subsequent

surveillance

Excessive pitting observed in some tendon wires of unit 2 during installation,

cause was outdoor storage in conjunction with high precipitation and inadequate

protection, defective tendons replaced

Zion I 50-295

Crystal River 3 50-30Z

1973

1978

1981

1972

1974 Twenty-eight-day concrete strength was low due to failure of cement to meet specLfications; design review revealed strength attained to be adequate; cement inspection Increased

197ba Dome delaminated over 32-a-diam (05-ft) area due to low concrete properties,

radial tension due to prestressing, and biaxial failure criterion; upper delamLnated section removed, additional rebars provided, concrete replaced, dome retenstoned, and structural integrity test conducted

Salem 2 50-311 1974 Incomplete concrete pour near equipment hatch due to use of wrung concrete mix,

voids repaired with high-strength nonshrink grout

Rancho Seco Concrete surface temperature >66'C (150'F) during initial power escalation

Cook 1/2

Calvert Cliffs 1/2

Three Mile Island 2

50-312 1975 1974

50-315/316 1975/1978 1974 Cracking in spent fuel pit wall and slabs framing into pit walls, cause was thermal

expansion and hydrostatic pressure, no threat to structural integrity

50-317/318 1975/1977 1 971/1 9 72a Eleven of top bearing plates of units 1I and 2 depressed into concrete because of

voids; 190 plates of each containment exhibited voids upon inspection; tendons

detensioned, plates grouted and tendons retensioned

0

50-320 1978 1974

1975

197b

Four of six sets of compression cylinders had low f because of mishandling and in- c ventory control at cement silo, 90-d strengths were acceptable and concrete inplace determined to have adequate strength; cement storage and sampling techniques

improved

Void 7 cm (3 in.) high 1.8 (6 t) wide x 0.9 to 1.5 m (3 to ft) deep occurred in south exterior wall of fuel-handling building, cause was improper placement, void determined not to be a threat to structural or shielding effectiveness;

void refilled

Void 0.9 to 1.2 (3 to 4 it) into concrete 0.4 a (1.5 t) high by 1.8 to 2.4 a

(6 to 8 fit) wide In north exterior wall of fuel transfer canal, void repaired, no

structural or shielding effectiveness threat

Cracks in concrete wall around base plate

Concrete in pedestal for several recirculation line snubbers exhibited spalling

and cracking due to design deviation, 2.5-cm (1-in.) plates with four wedge

anchors installed on top of existing plates

Unconsolidated and honeycombed areas in first lift of reactor support pedestal,

voids repaired after determining that they were not a threat to structural integrity, placement procedures mproved

Hatch I 50-321 1975

Uco

1981

1981

Shureham 50-322 1974
-P -

Year of Year of

Reactor plant Docket No. commercial of Summary description

operation occurrence

Brunswick 1/2 50-324/325 1977/1975 1974

Sequoyah 2

Midland 2

Duane Arnold

Fitzpatrick

Beaver Valley I

St. Lucie I

North Anna 2

Fermi 2

Davis Besse I

Farley I

San Onofre 3

50-328 1982

50-330 Cd

50-331 1974

50-333 1975

50-334 1977

50-335 1976

50-339 1980

50-341/342 1985

1978

1975

1977

1974

1973

1982

1974

1978

Voids occurred behind liter during construction of suppression chamber'grout

injected nto voids through holes drilled in liner, some grout in unit I did

not harden but was left'in place to provide limited resistance

Concrete in outer 2.5 to 5 c (I to 2 in.) of unit 2 shield building was understrength because of exposure to freezing temperatures at early concrete age,

determined not to affect shield building capability

Rebar spacing defictencies. in reactor containment buflding, determined error not

significant enough to affect safety

Leaking water pipe in exterior wall caused bulging of liner plate up to 0.6 

(2 ft) inwards over an area of about 195'm2 (2100 ft2) producing concrete

spalling of 7.5 to 25.4 cm (3 to 10 in.) deep, bulged liner plate and concrete

removed -

Hairline cracks'in floor under torus, cracks permitted to self heal

Horitzontal crack from hairline to 0.9 c (3/8 in.) wide in reactor pedestal

extending into concrete 0.2 to 0.7 (9 to 30 in.), cause holieved to be welding

procedure causing tension; structural tintegrity of pedestal not impaired, crack

sealed by epoxy njection

Void -0.9 (37 n.) long x 0.9 (3 ft) deep in outer containment wall n concrete

ring around equipment hatch,,no threat to structural ntegrity, void repaired

Concrete spalled because of scaffolding fire in annulus between containment vessel

and shield building, area affected -3.4 m (11 ft) x 0.6 (2 ft) x 2.5 cm'(1 n.),'

temperature reached 148 to 1770C (300 to-350*F) inflicting-only superficial

damage, spalled area replaced

Hairline crack -1 ma (1/24 in.) wide by I m (39 in.) long in east wall of reactor

containment refueling canal near embedded steel plate, crack'repaired by grouting

and column added to support-platform girder ' . - --

I-I-nI1974 Cracks >1.6 mm (1/16 In.) in containment floor slab occurred around neutron shield

tank anchorzbolts,following pressure testing of seal chambers, cause was

inadvertent pressurization,- cores showed cracks extended into concrete vertically,

cracks no tructural threat, routed and sealed to prevent fluid penetration

1972 Cracks <0.8 inn (1/32 in.) wide in basement floor slab permitted groundwater to seep

into building, cracks caused by shrinkage, cracks repaired by pressure grouting

after determining that they were no threat to structural Integrity

1984 Voids detected around one of auxiliary building watertight doors,'defective

concrete removed by chipping and area-grouted, other doors Inspected

1982 Two concrete expansion anchors and upper part of base plate' pulled from wall -cm

(3/8 in.) because of improper installation, anchors replaced and torque checked

1985 Cracks-detected in six containment tendon anchors during refueling outage

1980 Portions of unit I walls had areas where vertical reinforcing and grout were missing, corrective action taken

1983 Tendon liftoff force in excess of maximum value listed in technical specifications,

cause was lower relaxation rate than expected, no threat to structural integrity

50-346

50-348

50-362

1977

1977

1984
Year of Year of

Reactor plant Docket No. commercial of - Summary description

operation occurrence

Farley 2 50-364 1981 1985a Three anchor heads on bottom ends of vertical tendons failed and 18 cracked with

several tendon wires fractured, occurred about 8 years after tensioning, cause

attributed to hydrogen stress cracking, all tendons and anchor heads from same

heat inspected with no further problems noted, 20 tendons replaced

Hatch 2 50-366 1979 1979 Approximately 101 failures occurred during testing of 183 anchor bolts because of

improper installation, failed bolts replaced with wedge anchors

1982 Main steam pipe hangers had significant concrete spelling around embedded plate

with concrete missing -5 cm (2 in.) adjacent to plate, cause was defective

concrete material or faulty placement, plate redesigned

HcGuire I

La Salle 1/2

Waterford 3

50-369 1981 1976

50-373/374 1982/1984 1976

50-382 1984 1976

1976

1977

1977

1977

1984

Two buttonheads failed during stressing of CROW missile shield holddown tendons at

underside of bottom plate and two wires failed in another tendon near base anchor,

additional failed wires found during checking, cause was excessive corrosion,

design modified to replace tendons with 3.5-cm-diam (1-3/8-in.) threaded rods that

were grouted into place

Low concrete strength at 90 d, in-place strength determined acceptable from cores

and cement contents for future pours increased, strength low n only small percent

of pours so did not threaten structural integrity

Improper concrete placing sequences used in foundation met forming a cold oint and

not achieving stepped bedding planes, core drilling revealed fine cracks and

honeycombed areas, defective concrete removed and replaced, supervision and inspection ncreased.

Improper placement of concrete in reactor auxiliary building nterior wall resulted

in honeycombed areas, voids, and cold joints; unsound concrete removed and

repaired

Crane boom fell during construction on common foundation structure wall causing

concrete cracking and spalling over area 0.3 m (12 in.) x 10 cm (4 in.) x 2.5 cm

(I in.), rebars and concrete removed and replaced over entire height of damaged

area for a length of 9.5 m (31 t)

Low concrete compressive strength in 4.2 3 (5.5 yd3) of concrete in wall

contiguous with portion of condensate storage pool wall and wall of refueling

water pool

Low concrete strengthIn reactor auxiliary building slab, cores yielded satisfactory strength, amount of sand in future mixes increased as well as mixing

requirements

Spelled concrete observed in corbel exposing and displacing rebars and cracking

in plane of anchor bolts, no loss of structural support, area repaired

Coarse aggregate with excessive fines used because of quality control deficiency,

concrete strength exceeded requirements so structural integrity not affected,

aggregate material for future batches replaced

Voids located behind ither plate of reactor containment building wall, windows cut

in liner revealed voids up to 22 cm (8.5 n.) deep, cause was use of low slump

concrete with nsufficient compaction, voids chipped and cleaned to sound

concrete, filled with nonshrink grout and liner repaired with all welds leak

tested

Susquehanna 1/2 50-387/388 1983/1984 1976

Summer I 50-395 1984 1976
Year of Year of

Reactor plant Docket No. commercial of Summary description

operation occurrence

Summer I (continued) 1977 Excessive heat from welding caused liner attached to concrete on inside face of

concrete primary shield wail cavity to buckle and fail stud anchors and crack

concrete, liner and concrete to depth of 15 cm (6 n.) removed, new liner plate

welded n place and pace filled with high-strength grout

Hanford 2

Catawba 2

Grand Gulf 1/2

Bellefonte 1/2

Seabrook 1/2

Commanche Peak 1/2

Byron l

50-397

50-414

1966

Ucc

50-416/417 1984/UC0

50-438/439 UCO

50-443/444 UCC/Cd

50-445/446 UCC

50-454i 1984

1973 Deficiency in vertical cadveld splice sleeves n eactor building mat

1976 Cement used in reactor building base slab had been contaminated by fertilizer,

7-d strengths exceeded 28-d design values, cement feed transferred to another silo

1975 Seien of 19 cylinders for control building base slab concrete did-not meet 28-d

design strength, 90-d values vere acceptable ',

1976 Voids found beneath drywell wall embed and shear key because of too stiff'a

concrete ix; holes drilled through embed and used to fill voids with highstrength grout; voids below shear key repaired by removing central'portion of

plate, chipping to good concrete, adding rebars, replacing concrete and liner, and

leak testing liner - ' -

1984 Expansion shell anchor failures occurred in control building concrete because of

low surfi-ec''ncrete' strength,'anchors replaced by more deeply embedded bolts or

grouted anchors 

19 7 6 a Eight rock 'anchor heads failed durtng'constructton because of possible stress

corrosion cracking,-anchor heads replaced with cleaner steel

1983 Cracking occurred,in walls at end of stiffening slabs separating pump cells In

category I service water and circulating water pumphouse, cause was shrinkage

and;temperature variations, stiffening slabs were modified

1975 Cold jolnt formed n reactor nat, concrete removed, rebars exposed and new joint

poured ' ' ' ' .

1976 Volds lO'to'16'm(3/Sto'5/8 in.) found under'sump plates in concrete base mat,

voids filed with-neat cement grout using holes drilled through plates

1976 Concrete not properly compacted around one' of valve isolation embeds forming a

void, faulty,'naterial removed by-chipping and replaced by moita-or' concrete

1976 Inadequate concrete compaction under containment wall for 58 (190 ft) at 1.8 to

2.1 n-(6 to 7 ft) below top of mt, 3.7 by 6.1 (12 by 20 ft) area south of

reactor pit, 18 by 37 (6 by 12 ft) area south of north sump and 1.2 by 1.8 m

(4 by 6 ft) area north of. northasump; core holes drilled for-inspection n. 

conjunction with analytical evaluations'revealed',base1mat was adequate for all.,

loading'conditions; cores filledwith mortar and nterconnecting voids grouted

1976 Excessivemortar used in concrete placement in preparing joint at reactor cavity

wall, not determined' to be structurally detrimental ; -

1976 Fresh concrete placed in area'of standing water, because concrete forced water

ahead of placeeenttlt was not considered detrimental, excess water removed

1976 Hardened-concrete observed splattered on rebars, extent of occufrence considered

minor with-bond reduction nsignificant' ' '

1 9 7 9 a Four anchor head failures occurred in first year after stressing, cause was use of

vanadium grain refinement'process n conjunction with temperatures not high enough
Year of Year of

Reactor plant. Docket No. commercial of Summary description

operation occurrence

e * C z t r ~~~~_ _ _- - _ _ _ _ _ _ . _ - _ . . _ - _ j . ___ 

Clinton Sv-461 UC-. I984 EMbed plate on Outside o dryv ll all pulled trom concrete because ot aSlure o0

several Nelson studs occurring as result of weld shrinkage, concrete excavated

along plate edges, ebed plate redesigned and grout placed nto area'vhere

concrete ma .removed -

Wolf Creek

Callaway I

South Texas 1/2

50-482

50-483

1985

1985

1978 Low concrete strength n reactor building base at with some 90-d values below 28-d

values, 7in-situstrength tests Indicate concrete exceeded design values and low

90-4 strengths were due to testing conditions

1978 Voids up'to 1.8 (6 ft) wide and through-wall thickness occurred under equipment

and personnel hatches in reactor containment'building, voids repaired and quality

assurance'progran updated '

1977 Nineteen randomly located areas of honeycombing extending to bottom layers of

rebar n reactor building base mat in annular area of tendon access area, cause

was use of low.slump concrete in congested area, defective material removed from

33.of 172,tendon. trumplatec.and voLds repaired: '

50-498/499 UCC 1977

1977

1978

1978

1979

1983

Crack in-fuel'handLLng building wall due to shrinkage, no structural significance

Rebarsa'lproperly'located in buttress'region of unit'l containment, detailed analysis of s-built condition determined that no safety hazard to public occurred

Unconsolidated areas occurred in bottom surface of concrete slab in south unit I

fuel-hmndling building; material removed by chipping to expose rebars, surface

was epoxy-sealed followed by epoxy njection and a combination of dry packing,

shotcreting, and epoxy njection 1 - ' 

Voids occurred behind liner plate of unit I reactor containment building exterior

wall because of planning deficiencies, long pour time and several pump breakdowns;

sounding nd fiberoptic exam through holes drilled in liner plate were used to

determine extent, areasa were repaired by grout injection

Voids were detected n 12 areas behind liner plate of reactor containment building

exterior wall with cause being attributed to temporary weldments, normal concrete

settlement/shrinkage, and liner movement; construction and quality control procedures strengthened

Rust and pitting were observed on tendons for units 1 and 2 while in storage at

fabricating plant, cause was delayed and Improper corrosion nhibitor application

and storage in a facility without temperature and humidity control; detailed exam

of 14 tendons revealed pitting up to >0.4 us (15 ails) but strength and ductility

exceeded lmts;dsmged tendons were replaced and controlled storage conditions

utilized with properly applied corrosion inhibitor

Honeycombing around vertical tendon sheath blockouts with most voids at buttress/

shell nterface above last dome hoop tendon, condition was localized so area

repaired

U'

Palo Verde 2/3 50-529/530 UCo 1984
-

Year of Year of

Reactor plant Docket N*o. commercial of Sumary description

operation occurrence

Marble Hill 50-546 Cd- 1979 High concrete pour rate may have bowed liner

1979 A 0.3-u-deep (12-in.) void extending 6.1 x 1.4 (20 x 4.5 ft) in axial direction

in base slab for auxiliary building, void repaired by shotcrete njection

1979a Numerous surface defects (4000) and inadequate patching resulting from poor

concrete compaction and improperly prepared construction joints; breakdown in

quality control and construction management attribute as cause; internal concrete

inspection revealed it to be of high quality with higher than required strength;

patches removed and replaced using good construction practices; providing good

workmanship is used in repair and procedures followed, consultants determined

structural integrity and shielding requirements should be met

aescribed in re detail in Sect. 3.3.2.

bDescribed in more detail in Sect. 3.2.2.

CUnder construction.

dCancelled or indefinitely deferred.

ILn

LA
'157

NUREG/CR-4652

ORNL/TM-10059

*Dist.-Category RV

Internal Distribution

1 .

2.

3.-

4-6.

7.

8.

9.

10.

11.

12.

13.

14.

J. M. Corum

W. G. Craddick

L. R. Dole

D. M. EissenbergJ. A. Getsi,-

W. L. Greenstreet

H. D. Haynes ;

J. E. Jones Jr.

P. R. Kasten

R. C. Kryter

A. P. Malinauskas

G. C. Murphy

15.

16-19.

-20.

21.

I . 22.

23.

'" 24.

25.

-26."

27.

28-29.

30.

F.'R. Mynatt

D. J. Naus

C. B. Oland, '

C.'E. Pugh

G.' C.: Robinson

H. E. Trammell

D. B. Trauger

ORNL Patent Office'

Central Research Library

Document Reference Section

Laboratory Records Department

Laboratory Records (RC) -

External Distribution

31. Gunter Arndt, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

32. Hans Ashar, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

33. A. B. Johnson, Pacific Northwest Laboratory, P.O. Box 999,

Richland, WA 99352

34. Don Lamken, Commonwealth Edison Company, One First National

Plaza, P.O. Box 767 (Room 35FNW), Chicago, IL 60690

35. Alexander Marion, Rm. 920, Baltimore Gas and Electric Co., P.O.

Box 1475, Gas and Electric Building, Baltimore, MD 21203

36. J. W. McElroy, Philadelphia Electric Co., P.O. Box 8699,

Philadelphia, PA 19101

37. C. Michelson, ACRS, 20 Argonne Plaza, Suite 365, Oak Ridge, TN

37830

38. Jerry E. Rhoads, Washington Public Power System, 3000 George

Washington Way, P.O. Box 968, Richland, WA 99352

39. R. L. Simard, Institute for Nuclear Power Operations, 1100

Circle 75 Parkway, Atlanta, GA 30339

40. George Sliter, Nuclear Engineering and Operations Department,

Electric Power Research Institute, 3412 Hillview Avenue, P.O.

Box 10412, Palo Alto, CA 94303

41. D. F. Sullivan, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 5650 Nicholson Lane, Rockville, .MD 20852

42. J. H. Taylor, Engineering and Risk Assessment Division, Department of Nuclear Energy, Brookhaven National Laboratory, Upton,

NY 11973
158

43. Jim E. Thomas, Duke Power Company, P.O. Box 33189, 422 South

Church Street, Charlotte, NC 28242

44. Charles Turk, Arkansas Power and Light Co., P.O. Box 551,

Little Rock, AR 72203

45. Stanley Urbanowski, Yankee Atomic Electric Co., 1671 Worcester

Road, Farmington, MA 01701

46. J. P. Vora, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

47. Don R. Woodlan, Licensing Engineer, Texas Utilities Generating

Co., 400 North Olive Street, LB81, Dallas, TX 75201

48. Office of Assistant Manager for Energy Research and Development, Department of Energy, Oak Ridge Operations Office, Oak

Ridge, TN 37831

49-50. Technical Information Center, Department of Energy, Oak Ridge,

TN 37831

51-380. Given distribution as shown in NRC category RV (10 - NTIS)
-

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I REPORT NUMBER Ass.gfnedy TIDC. eadd VOl ho. 11`41

12.841

3201102. NUREG/CR-4652 3R011202. BIBLIOGRAPHIC DATA SHEET ORNL/TM-10059

SEE INSTRUCTIONS ON THE REVERSE

2 TITLE AND SUBTITLE 3 LEAVE BLANK

Aging of Concrete Components and Its Significance

Relative to Life Extension of Nuclear Power Plants

S AUTHORCSI

D. J. Naus

4 DATE REPORT COMPLET 

MONTH I ER

August 1986

6 DATE REPORT ISSUED

MONTH I YEAR

September I 1986 7. PERFORU;NG ORANIZAI 0h NAMEAO MAILINGADDRESS ico.ZoDCde

Oak Ridge National Laboratory

P.O. Box X

Oak Ridge, Tennessee 37831

8 ROJECT:TASK,WORK UlNIT NUMBHE"

9 FIN OR GRANI NUMBER

B0828

.4- 10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS IMcI.,wZpeCod.I

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Division of Engineering Technology

Washington, DC 20555

Ila. TYPE OF REPORT

Topical

.PERIOD COVERED I/clsgw{* r .

. _ _. .__ ._,, ,,_, .A_*- _ 8--_-_, .

12 SU'PLEMENTARY NOTES

13 ABSTRACT 1200 wades 0t ,!es/

The objective of the study was to: (1) expand upon the work which was initiated in

the first two EPRI studies relative to longevity and life extension considerations of

safety-related concrete components in LWR facilities, and (2) develop background which

will logically lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based materials and components. These objectives are consistent with NPAR Program Goals:, (1) to identify

and characterize aging and service wear effects which, if unchecked, could cause

degradation of structures, components, and systems and thereby impair plant safety;

(2) to identify methods of inspection, surveillance and monitoring, or of evaluating

residual life of structures, components, and systems, which will assure timely detection of significant aging effects prior to loss of safety function; and (3) to evaluate

the effectiveness of storage, maintenance, repair and replacement practices in mitigating the rate and extent of degradation caused by aging and service wear.

tB. M. Morris and J. P. Vora, "Nuclear Plant Aging Research (NPAR) Program Plan,"

NUREG-1144, Division of Engineering Technology, Office of Nuclear Regulatory Research,

U.S. Nuclear Regulatory Commission, Washington, DC (July 1985).

14 DOCUMENT ANALYSIS- KEYWORDS:DESCRiPTORS

Aging

Concrete

Evaluation Methodology

In-service Inspection

a IOETIFIERS OPEN ENDED TERMS

Repair

15 AVAILABILITY

STATEMENT

Unlimited

16 SECUAITYCLASSIFICATION

Unclassified

I TA4s .Om~

Unclassified

17 NUMBER OF PAGES

19 PRIPS

_ .

*U.S. GOVERNMENT PRINTING OFFICE 986-631056140043