ORNL-IAEA-(Concrete Structure Aging Overview) Slides_2012.pdf

1 Managed by UT-Battelle for the U.S. Department of Energy 1 Managed by UT-Battelle for the U.S. Department of Energy

An Overview of NPP Safety-Related

Concrete Structures and Activities at ORNL

In Support of Continuing their Service

Dan Naus

Materials Science & Technology Div.

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37831-6069

First Consultancy on Assessment & Management

of Concrete Containment Buildings

IAEA Vienna

29 May to 01 June 2012


Presentation topics

Nuclear power plant concrete structures Importance

Materials of construction

Factors that can produce degradation

In-service inspection and examination requirements

License renewal

Operating experience

Overview of concrete research at ORNL

U.S. Nuclear Regulatory Commission - sponsored activities

U.S. Department of Energy - sponsored activities


All NPPs contain concrete structures whose

performance and function are necessary to protect the

safety of plant operating personnel and general public

Concrete structures are essentially passive under normal operating conditions, but play a key role in mitigating the impact of extreme/abnormal operating and environmental events

Structural components are somewhat plant specific, may be difficult to inspect, and usually can not be replaced

Structures are subject to time-dependent changes that may impact their ability to withstand various demands from operation, the environment, and accident conditions

Excessive degradation can lead to failure

Failure often affects serviceability, not safety

As NPPs age, assurances need to be provided that the capacity of the safety-related systems to mitigate extreme events has not deteriorated unacceptably due to either aging or environmental effects


NPP safety-related concrete structures are

composed of several constituents that, in

concert, perform multiple functions

Concrete Mild Steel Reinforcement

Post-tensioning tendons Steel Liner Plate

Construction of

PC Containment


Summary of common causes of defects in

concrete members

Cracking Due to

Alkali-Silica Reaction

Chloride Ingress

Unsuitable Materials Improper Workmanship Environmental Exposure Structural

Aggregate

unsound or reactive

contaminated

Cement

wrong type

manufacturing error

contaminated

Admixture

wrong kind

contaminated

Water

organic contaminants

chemical contaminants

dirty

Reinforcement

wrong kind

incorrect size

Faulty Design

Incorrect Concrete Mix

low cement content

high water content

incorrect admixture

dose

batching errors

High Slump

Unsuitable Formwork/Shoring

Misplaced Reinforcement

Handling/Placing Concrete

segregation

careless placing

inadequate or over vibration

poor finishing

Incomplete Curing

Concrete

Chemical Attack

efflorescense or leaching

sulfates

acids or bases

delayed ettringite formation

alkali-aggregate reactions

Physical Attack

salt crystallization

freezing and thawing

thermal exposure/thermal cycling

abrasion/erosion/cavitation

irradiation

fatigue or vibration

biological attack

Steel Reinforcement

carbonation, chlorides and stray currents

Loads Exceed Design

Accident

Settlement

Earthquake

Carbonation


Concrete containments are metal-lined RC

pressure-retaining structures that in

some cases may be post-tensioned

Initially ACI Standard 318, “Building Code Requirements for Reinforced Concrete,” used as the design basis

Additional criteria (e.g., loads and load combinations) for design of seismic Category I structures were developed because ACI 318 was not considered adequate and did not cover the entire spectrum of design requirements

Current design rules provided in ASME Section III, Division 2 (ACI 359) with supplemental load criteria provided in Sections 3.8.1 and 3.8.3 of the NRC Standard Review Plan

Design and construction requirements for non-containment-related safety-related concrete structures is contained in ACI 349


Typical design parameters for selected

PWR plants

Type Material Ref. Plant Int. Diam

(m)

Free Vol.

(103 m3)

Des. P

(kPa)

Des. Leak Rate

(% Vol./day)

Large Dry RC Hemi. Dome Indian Pt. 3 41 74 324 0.1

Large Dry St. Cyl. Hemi Dome Davis Besse 40 81 276 0.5

Large Dry PC Shallow Dome Zion 43 81 324 0.1

Large Dry PC Hemi. Dome Trojan 38 57 414 0.2

Ice Condenser St. Cyl. Hemi. Dome Sequoyah 32 -- 74 0.5

Subatmospheric RC Hemi. Dome Surry 38 51 310 0.1

Large Dry –

Diablo Canyon

Subatmospheric –

North Anna 1

Ice Condenser –

Cook 1


Typical design parameters for selected

BWR plants

Drywell Wetwell

Type Material

Ref. Plant Des.

Temp

(˚C)

Free Vol.

(103 m3)

Des.T

emp

(˚C)

Free Vol.

(103 m3)

Des.

Pressure

(kPa)

Des. Leak Rate

(% Vol/day)

Pre MK Steel Sphere Big Rock Pt. 113 33 --- --- 186 0.5

MKI Steel Peach Bottom 138 5 138 4 386 0.5

MKII Rein. Concrete Limerick 138 7 138 4 386 0.5

MKIII Rein. Concrete Grand Gulf 166 8 85 36 103 0.35

MKI –

Peach Bottom MKIII –

Grand Gulf

MKII –

Limerick


Current in-service inspection and

examination requirements

10 CFR Part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors”

10 CFR Part 50.55a, “Codes and Standards” – Adoption of ASME Section XI, Subsections IWE and IWL

10 CFR Part 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants”

Vogtle NPP


Appendix J to 10 CFR 50 sets testing requirements

for preoperational and periodic verification of leak-

tight integrity of containment

Addresses overall leakage rate (ILRT - Type A), local leaks at penetrations (Type B), and isolation valve leakage rates (Type C)

Option A (fully deterministic) requires that after preoperational leakage rate tests a set of three ILRTs be performed at approximate equal intervals during each ten-year service period

Option B does not provide a quantitative requirement for scheduling the Type A tests

NEI 94-01, “Industry Guidance for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J”

RG 1.163, “Performance-Based Containment Leak-Test Program”

General visual inspection of accessible interior and exterior accessible surfaces of containment and components is required prior to a Type A test to identify evidence of structural deterioration that may affect structural integrity or leak-tightness


In August 1996, the NRC amended its

regulation 10 CFR Part 50.55a

General visual inspections under Appendix J were not done in a consistent manner

No formal procedure for documenting the degradations and implementing corrective actions

Degradation occurrences exhibited a trend to be increasing

Endorsed Subsections IWE and IWL of Section XI of the ASME B&PVC

Subsection IWE provides the requirements for in-service inspection, repair, and replacement of Class MC pressure-retaining components and their integral attachments, and metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments

Subsection IWL provides the requirements for pre-service examination, in-service inspection, and repair of the reinforced concrete and post-tensioning systems of Class CC components


ASME in-service inspection and repair

guidelines flow diagram


ACI 349.3R provides visual-based approach

for assistance in classification and

resolution of inspection findings

• Acceptance without further evaluation

• Acceptance after review

• Conditions requiring further evaluation


“Maintenance Rule (10 CFR 50.65)” issued

by NRC to reduce likelihood of failures due

to degradation

Owners must monitor performance/conditions of structures, systems, and components (SSCs) against owner-established goals to provide assurances that functions are being fulfilled

Owners must take timely and appropriate corrective action when performance or condition of SSCs does not conform to established goals

NUMARC 93-01 (Rev. 2), “Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants”

R.G. 1.160 (Rev. 2), “Monitoring the Effectiveness of Maintenance at Nuclear Plants”

IP 62002, “Inspection of Structures, Passive Components, and Civil Engineering Features at Nuclear Power Plants”

IP 62003, “Inspection of Steel and Concrete Containment Structures at Nuclear Power Plants”

IP 62706, “Maintenance Rule”


“Requirements for renewal of operating

licenses for nuclear power plants (10 CFR

Part 54)” Focuses on managing the adverse effects of aging rather than

identification of all aging mechanisms

Ensures that SSCs will continue to perform intended functions in period of extended operation (POE)

Addresses passive, long-lived components because regulatory process and existing licensee programs may not adequately manage the detrimental effects of aging during the PEO

Key principles are that regulatory process (continued into the PEO) is adequate to ensure that current licensing basis of all currently operating plants provides an acceptable level of safety, with the possible detrimental effects of aging on certain SSCs and possibly a few other issues related to safety only during the PEO; and each plant’s current licensing basis is required to be maintained during the renewal term

Environmental aspects addressed through “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions (10 CFR Part 51)”


License renewal methodology and key

documentation

Standard Format and Content for Application to Renew NPP Operating Licenses (Regulatory Guide 1.188)

Standard Review Plan for Review of License Renewal Applications for NPPs (SRP-LR, NUREG-1800)

Generic Aging Lessons Learned (GALL) Report (NUREG-1801)

Catalogs plant structures and components

Identifies specific material they are composed of and their associated environments

Lists aging effects

Documents staff’s evaluation of generic aging programs that can mitigate or manage these aging effects


Status of license renewal applications and

locations of applied-for new NPPs

U.S. Commercial NPP Operating Licenses –

Issued by Year and License Renewal Status

Locations of Applied-for New NPPs

(status as of June 30, 2011)

Ten plants have entered the operating period beyond 40 years


Operating experience indicates that concrete

structures have a history of reliability and

durability, but there have been occurrences of

degradation

Containment Dome

Delamination Repair Water Intake Structure

Rebar Corrosion

Concrete Wall

Water Infiltration

Concrete Cracking

Outside Containment Wall

Anchor Head Failure

Corrosion of

Grease Cap Spent Fuel Pool

Leakage

Grease Leakage

Outside Containment Wall


AMP audits performed as part of the license

renewal process have identified areas of interest –

concrete inspection

Licensees conduct periodic visual inspection of concrete structures, however the inspection criteria for visual inspection vary from plant to plant

The GALL report recommends use of ACI 349.3R for frequency and quantitative inspection criteria

NRC issued Information Notice 2010-14 to inform licensees about recent operating experience related to containment concrete surface condition examination frequency and acceptance criteria

NRC recently issued Information Notice 2011-20 to inform licensees about potential for concrete degradation due to alkali-silica reaction

New revision of GALL report provides additional and specific guidelines for inspection

Crack Comparator



renewal process have Identified areas of interest –

spent fuel pool leakage

Spent fuel pool leakage identified at several nuclear power plants.

Leakage occurs thru the seam or plug welds of the stainless steel liner plate.

Leak chase channels designed to collect water from weld seams blocked by boron crystals.

Licensee’s have been successful in cleaning the leak chase channels.

Studies conducted by the industry concluded that low concentrations of borated water does not affect the concrete and rebar.

NRC conducting confirmatory study and testing to determine the effect of borated water on concrete and rebar.

Cross section of a leak chase system for a

PWR plant and postulated path for

leakage occurrence



renewal process have identified areas of interest –

reactor refueling cavity leakage

Leakage at several PWR and BWR plants

Leakage occurs when the reactor cavity is flooded during plant outage

Leakage normally thru the welds in the liner plates

Volume varies between 2-100 gpd

Leakage travels thru concrete and can corrode liner plate

Difficult to pin point leakage source

Licensees have tried different types of coatings on stainless steel liner plate to stop leakage

Detail of drywell-reactor cavity seal area

and identification of potential leak path

Source: ADAMS ML110070342


As operating experience increases, we may

see “surprises” – backside liner corrosion

Four plants have identified liner

corrosion perforation btw 1999 and 2009

Conceptual Drawing

Illustrating Liner Repair

Moisture Barrier

Paint Blister After Corrosion

Product Removal

Embedded Wood

SGR Replacement

D.S. Dunn, A.L. Pulvirenti, and M.A. Hiser, “Containment Liner Corrosion Experience Summary Technical Letter

Report,” USNRC, August 2, 2011 (ADAMS ML112070867).

J.P. Petti, D.J. Naus, A. Sagüés, R.E. Weyers, B.A. Erler, and N.S. Berke, ”Nuclear Containment Steel Liner Corrosion

Workshop: Final Summary and Recommendations Report,” SAND2010-8718, July 2011 (ADAMS ML112150012).


As operating experience increases, we

may see “surprises” – Crystal River 3

containment delamination

Steam Generator Replacement Opening

• At liner – 23’ 6“ by 24’ 9”

• At concrete opening – 25’ 0” by 27’ 0”

Ref. - ADAMS Accession Number ML102861026


Root cause analysis – investigation

approach

Complex investigation conducted that considered 75 potential failure modes

Containment design and analysis

Concrete construction

Use of concrete materials

Shrinkage, creep, and settlement

Chemically- or environmentally-induced distress

Concrete-tendon-liner interactions

SGR containment cutting

Operational events

External events

Non-destructive testing of containment wall surfaces

Impulse-response (IR) and ground-penetrating radar (GPR) with over 8,000 IR readings taken

Addressed all accessible surfaces


Root cause analysis – investigation

approach (cont.)

Concrete cores

Over 150 cores obtained

Ranged from 1” to 8” diameter by 6” to 32” long

Validated IR data and boroscopic investigations

Laboratory testing

Petrographic examination

Modulus of elasticity and Poisson’s ratio

Compressive strength, splitting-tensile strength, and direct-tensile strength

Fracture energy

SEM examination of micro-cracking

Density, absorption, and voids


Root cause analysis and post-repair test

plans

Root Cause Summary

Delamination occurred during the outage

Detensioning scope and sequence resulted in redistribution of stresses that exceeded tensile capacity

Result could not have been predicted based on existing information and available models at time of delamination

Post-Repair Testing

Structural integrity test at 1.15 times peak design pressure (63.3 psig)

Integrated leak rate test required per ASME Section XI

Repair of Delaminated Area



see “surprises” – “hairline” cracks in Davis-

Besse shield building architectural concrete

Free-standing

SCV (1.5”)

Annulus

Space

(4.5 ‘)

Free-standing

RC Shield

Building (2.5’)

Note: 2002 cross-section during RPV closure head replacement.

Not current opening. Architectural concrete not visible.

Tight cracking near outer rebar mat Flute region

Top 20’ of SB wall outside flute shoulder regions

Two regions adj. MSL penetrations

Impulse-response testing plus obtaining concrete cores

Root cause determination w additional monitoring

Shoulder

Flute



see “surprises” – alkali-aggregate reactions

at Seabrook

(a) Mechanism

(b) Resulting gel that causes

expansion and cracking

(c) Polished section showing

internal cracking

Cracking observed in

exterior walls of control

building electrical tunnel

(a) (b)

(c)


ORNL has conducted concrete research

in support of NRC and DOE programs since the

mid-1960’s

U.S. Nuclear Regulatory Commission

• Regulatory Guide Evaluations • High-Temperature Gas-Cooled Reactor Studies • Nuclear Power Plant Aging Studies • Structural Aging Program • Inspection of Aged/Degraded Containments • Thermal Effects on Concrete • Special Investigative Team for Crystal River 3 • Civil/Structural Review of License Renewal Applications • Technical Assistance for License Renewal • Irradiation Effects on Concrete

U.S. Department of Energy

• Prestressed Concrete in Nuclear Pressure Vessels • Gas-Cooled Fast-Breeder Reactor • Clinch River Breeder Reactor • PCPVs for Coal Gasifiers • LWR Sustainability Program


U.S. Nuclear Regulatory Commission

Programs


Structural Aging Programs – ORNL is helping to

provide evidence that NPP structures will

continue to meet requirements for protection of

public health and safety

The Structural Aging Program (1988-1996)

An Investigation of Tendon Sheathing Filler Migration Into Concrete (1997-1998)

The Inspection of Aged/Degraded Containments Program (1993-2001)

The Effect of Phosphate Ion on Concrete Program (2004-2006)

The Environmental Effects on Containments and Other NPP Structures (2004-2008)

The High Temperature Effects on Concrete Program (2007-2010)

Technical Assistance for License Renewal Related to Civil Structures (2009-2012)

The Radiation Effects on Concrete Program (2010-2012)


High Temperature Effects - Elevated

temperature concrete material property

data and information have been compiled

Mechanical Physical

Stress-strain

Poisson’s ratio

Modulus of elasticity

Compressive strength

Tensile strength

Shrinkage/creep

Concrete-steel bond

Fracture energy

Fracture toughness

Long-term exposure

Radiation shielding

Multiaxial loadings

Porosity/density

Coefficient of thermal expansion

Thermal conductivity

Thermal diffusivity

Specific heat

Heat ablation/erosion rate

Moisture diffusion/pore pressure

Simulated hot spots

• Radiation shielding concretes

• Codes and Standards

• Potential methods for assessment of concrete exposed

to elevated temperature

• Temperature-dependent properties of mild steel and

prestressing materials

• General behavior

• Properties

NUREG/CR-7031


Technical assistance for license renewal

related to civil structures

Review and evaluation of technical issues

Leakage from refueling cavity

Spent fuel pool leakage

Torus corrosion

Concrete safety-related structures

Case studies developed for leakage from refueling cavity, spent fuel pool leakage, and torus corrosion

Introduction

Field observations

Design characteristics

Corrective actions

Structural integrity assessments and test results

References Concrete Cracking in Wall

Spent Fuel Pool

Cross-Section of BWR MK I


Plant license renewal applications and

safety-evaluation reports provided primary

sources of information

NPP

RFC (11/7)*

SFP (12/9)*

Torus (7/7)*

Concrete (26/0)*

Beaver Valley 1 X

Browns Ferry 2 X

Browns Ferry 3 X

Brunswick 1 X

Cooper X

Crystal River 3 X X X

Davis-Besse X X X X

Diablo Canyon 1 X X

Diablo Canyon 2 X X

Duane Arnold X X

FitzPatrick X

Hope Creek X X X

Indian Point 2 X X

Kewaunee X X X X

Monticello X X

Nine Mile Point 1 X X

Nine Mile Point 2 X

Oconee X

Oyster Creek X X X

Peach Bottom 2 X

Peach Bottom 3 X

Prairie Island 1 X

Prairie Island 2 X

Palo Verde 1 X

Pilgrim X

Salem 1 X X X

Salem 2 X X X

Seabrook X X

Three Mile Island 1 X

Turkey Point 3 X

Vogtle 1 X

Vogtle 2 X X

*(# Events/# Case Studies) NUREG/CR-7111


Concrete degradation was primarily in form

of cracking/spalling/loss of material due to

aggressive chemical attack

Aging Mechanism Events

Hydrogen-induced SCC 1

Aggressive chemical attack 9

Leaching 7

Stress relaxation 1

Corrosion 5

Service-induced loads 5

Elevated temperature 1

Aging Effect Events

Loss prestressing 2

Cracking/spalling/loss

material

20

Increase in porosity and

permeability

3

Loss of strength 1


Irradiation Effects - Data and information

addressing irradiation effects on concrete

microstructure and performance are limited

H.K. Hilsdorf, ACI SP-55,

American Concrete Institute, 1978.

Literature

Sampling and

Testing


Under the USNRC Program data and

information addressing irradiation effects

on concrete microstructure and

performance are being assembled

Background Physics

Interaction mechanisms of irradiation and concrete constituents

Impact of interaction (attenuation, heat generation, and concrete degradation)

Concrete for radiation shielding Constituents

Guidance used for design of concrete radiation shields

Typical concrete used as radiation shields

Experimental results on impact of irradiation on concrete properties General behavior (separation of thermal and nuclear effects?)

Mechanical properties

Physical properties

Impact on durability (e.g., ASR acceleration)

Shielding effectiveness

Working with Professors Willam (University of Houston)

and Xi (University of Colorado)


Under the USNRC Program data and

information addressing irradiation effects

on concrete microstructure and

performance are being assembled (cont.)

Codes and standards that address concrete exposed to neutron and gamma radiation Current practice

Adequacy of exposure limits provided in codes and standards

Potential methods for assessment of concrete exposed to irradiation Visual assessment

Field-testing techniques

Laboratory techniques

Summary, conclusions and recommendations for further research Use of elevated temperature as a proxy for irradiation exposure

Obtaining and testing samples from biological shields and RPV supports of decommissioned NPPs or research reactors

Impact of neutron and gamma exposure on NPP structures after 40, 60, 80, and 100 years operation

Significance of nuclear heating associated with irradiation


U.S. Department of Energy

Programs


LWR Sustainability Program is addressing

candidate concrete research areas

identified in NRC/DOE workshop

Develop material property database to provide data and information on the variation of material properties under the influence of aging and environmental stressors

Utilize decommissioned plants to compile material property data for evaluation of construction quality, long-term performance and trending, characterization and evaluation of environmental effects, and assessment and validation of NDE and repair methods

Evaluation of long-term effects of elevated temperature and radiation

Develop improved damage models and acceptance criteria

Non-intrusive methods for inspection of thick-walled, heavily reinforced concrete structures and basemats

Global inspection methods for metallic pressure boundary components including inaccessible areas and backside of liners

Data on application and performance (e.g., durability) of repair materials and techniques

Utilization of structural reliability theory incorporating uncertainties to address time-dependent changes to structures to demonstrate that minimum accepted performance requirements are exceeded or to indicate on-going degradation to estimate end-of-life

Application of probabilistic modeling on component performance to provide risk-based criteria to evaluate how aging affects structural capacity


Activities under LWRS Program are in

support of continuing the service of NPPs

Nuclear Concrete Materials Database (NCMDB)

Data and information for input into NCMDB

Literature

Obtaining and testing samples

NPPs and research reactors (e.g., SRS, Hanford, Zion)

Decommissioned reactor containment – Barsebäck 1

Risk-informed basis to evaluate aging of NPP concrete structures

Plant management system for determining the condition and residual lifetime of concrete containments

H.K. Hilsdorf, ACI SP-55,

American Concrete Institute, 1978.


NCMDB is utilizing the database management

tools developed for the GEN IV materials

handbook

Information is provided in four volumes Volume 1 – Performance Data

Volume 2 – Supporting Data

Volume 3 – Baseline Data

Volume 4 – Organization and Revision Control Procedures

Materials are included as Chapters in Volumes 1, 2, and 3

Chapter 1 – Portland Cement Concretes Insulating

Structural lightweight

Normal weight

Heavyweight

Chapter 2 – Metallic Reinforcements Carbon steel bars

Stainless steel bars

Steel wires

Bar mats/wire fabric

Chapter 3 – Prestressing Tendons Carbon steel bars

Carbon steel wires

Strand

Nonmetallic materials

Chapter 4 – Structural Steels Carbon steels

Stainless steels

Chapter 5 – Rubbers

February 20, 2012

ORN/TM-2011/296

Phase I Complete and located on

Internal server at ORNL


Volume 1 of NCMDB provides

performance data (design values)


Volume 2 of NCMDB provides supporting

documentation

Information Sources

Reference 24: American Concrete Institute Journal. This information source provides constituent material

properties and test results for this material which is one material examined at the University of Wisconsin

under a long-term concrete study that began in 1910.

•

“Recommended Property” – Evaluation for Quality Level A

Requirement 1: Completeness of Material Description

The material description is complete and includes the mix proportions, constituent materials, plastic concrete

properties, and curing procedures.

•

Cement

vendor

Compressive Strength Test Results for Specimens Stored Outside, MPa (psi) at:

7 days 28 days 1 year 5 years 10 years 25 years 50 years

Medusa

(3M)

16.0

(2325)

22.9

(3315)

31.6

(4580)

46.7

(6780)

50.1

(7260)

51.0

(7400)

49

(7110)

Lehigh

(4M)

18.1

(2620)

23.8

(3455)

33.9

(4910)

47.8

(6930)

49.6

(7195)

52.1

(7555)

59.7

(8660)

• • • • • • • •

Average 17.7

(2570)

25.7

(3725)

34.0

(4930)

48.0

(6955)

49.9

(7240)

53.1

(7700)

54.4

(7890)

Test specimens were cast with each of four cements. Moist cured for 28 days, and then placed outside in

Madison, Wisconsin for long-term storage. Each value listed above is the average compressive strength

(Property Code 2013) from five test specimens (Reference 27).


Volume 3 of NCMDB provides material

data Property Code 2000 Constituent Material Information

Constituent Material Constituent Material Description or Property Property Value Property Code

Portland Cement

ASTM C 150, Type I Medusa (3M)

Fineness

Specific Gravity

•

110.0 m2/kg

3.12

•

2001

2101

2125

•

Fine Aggregate

Well-graded Janesville Sand (about 60% quartzite,

30% dolomite and 10% largely igneous material)

Fineness Modulus

•

2.9

•

2302

•

Coarse Aggregate

Janesville Gravel (~50% crushed material,

consisting of 75% dolomite, 20% quartz, and 5%

igneous material

Maximum Size

•

38 mm

•

2351

•

Water 2421

Property Code 2600 Plastic Concrete Properties

Plastic Concrete Property Property Value Property Code

Cement Content 369 kg/m3 622 lb/yd3 2601

• • • •


Volume 4 of NCMDB addresses organization

and revision control procedures

Handbook organization and updating information

Material code identification and description

Property code identification and description

Quality level criteria for data and values

References

Electronic data base description


NCMDB will include data and information

on aging as well as effects of pertinent

environmental stressors

Aging Elevated temperature

Irradiation

100 101 102 103 104 105

Age, days

0

0.5

1

2.5

2

1.5

Rela

tive

Co

mp

res

siv

e S

tren

gth


Currently working with Professor

Ellingwood (GIT) to provide a risk-informed

basis for evaluating performance of aging

NPP concrete structures

Provide appraisal of vulnerability of existing structures to intensities of natural and man-made hazards using recent research results on structural resistance and loads

Identify set of structures to be used as test beds to demonstrate the risk-informed condition assessment process

Identify major sources of aleatoric and epistemic uncertainties in engineering demand and capacity of the structures and develop probabilistic models of uncertainties

Develop risk-informed guidelines for evaluation of the critical structures identified above using structural reliability tools to model the uncertainties


Results will help provide improved criteria

to evaluate ability of existing structures to

achieve desired performance level when

subjected to uncertain demands

Which aging factors are significant for facility performance?

Has the original capacity degraded?

How would the structure respond to an event beyond the design envelope?

What is the remaining service life?

What uncertainties are significant and how should risk be managed?

Mechanics of degradation in service

Performance goals for new and existing construction

Reliability analysis tools

Supporting databases and models of uncertainty

Management of uncertainties and risk

Illustration of Application to

ISI/Maintenance Strategies

Structural Condition Assessment

Potential Applications of Results


ORNL participation in CONSAFESYS

Project is through DOE LWRS Program

Organization Country Funding Activity

Scanscot Technology* Sweden In-kind, cash Project team

Force Technology Denmark In-kind Project team

Peab Sweden In-kind Project team

Barsebäck NPP Sweden Decommissioned site Site activities

Oskarshamn NPP Sweden Site in operation, Cash Site activities

Oak Ridge National Laboratory U.S. Cash -

Imperial College U.K. In-kind Post Doc

Development Fund of the

Swedish Construction Industry

Sweden Cash -

Lund University Sweden - Ph D

*Project coordinator


System will utilize a combination of NDT

techniques and FE analysis to keep track

of present and predict future condition of

structures

Anticipated Accomplishments

Validation of NDT techniques at realistic circumstances

Development of the use and benefit of numerical models

Improved understanding of practical implementation at site in operation

Determination of the possibilities and limitations of seismic NDT methods at NPPs

Material testing database including core results

Barsebäck NPP Unit B1

Oskarshamn NPP Unit 3


CONSAFESYS status:

Measurements at site in operation: Oskarshamn Unit 3 2007: Finalized and reported

2010: Finalized, reported in September 2011

Measurements at decommissioned site: Barsebäck Unit 1 2001 – 2004: Finalized and reported

First phase of new testing: December 2011

Two test sessions planned for 2012

Numerical simulations of NDT testing Model of part of the Barsebäck containment wall

First analyses completed

Technical held workshop in October 2011