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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
2 Managed by UT-Battelle for the U.S. Department of Energy 2 Managed by UT-Battelle for the U.S. Department of Energy
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
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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
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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
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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
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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
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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
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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
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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
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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
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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
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ASME in-service inspection and repair
guidelines flow diagram
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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
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“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”
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“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)”
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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
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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
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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
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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
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AMP audits performed as part of the license
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
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AMP audits performed as part of the license
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
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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).
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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
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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
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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
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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
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As operating experience increases, we may
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
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As operating experience increases, we may
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)
29 Managed by UT-Battelle for the U.S. Department of Energy 29 Managed by UT-Battelle for the U.S. Department of Energy
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
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U.S. Nuclear Regulatory Commission
Programs
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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)
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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
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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
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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
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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
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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
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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)
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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
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U.S. Department of Energy
Programs
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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
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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.
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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
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Volume 1 of NCMDB provides
performance data (design values)
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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).
45 Managed by UT-Battelle for the U.S. Department of Energy 45 Managed by UT-Battelle for the U.S. Department of Energy
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
• • • •
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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
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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
48 Managed by UT-Battelle for the U.S. Department of Energy 48 Managed by UT-Battelle for the U.S. Department of Energy
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
49 Managed by UT-Battelle for the U.S. Department of Energy 49 Managed by UT-Battelle for the U.S. Department of Energy
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
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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
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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
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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