Information Notice 2011-20, NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011)

NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011)

ML19205A432
Person / Time
Site:
Issue date:	07/24/2019
From:	NRC/OGC
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
50-443 LA-2, ASLBP-17-953-02-LA-BD01, RAS 55108
Download: ML19205A432 (8)

UNITED STATES OF AMERICA

NUCLEAR REGULATORY COMMISSION

ATOMIC SAFETY AND LICENSING BOARD

In the Matter of

NEXTERA ENERGY SEABROOK, LLC

(Seabrook Station, Unit 1)

Docket No. 50-443-LA-2

ASLBP No. 17-953-02-LA-BD01

Hearing Exhibit

Exhibit Number:

Exhibit Title:

NRC060

NRC Information Notice 2011-20: Concrete Degradation by

Alkali-Silica Reaction (Nov. 18, 2011)

ML112241029 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001

November 18, 2011

NRC INFORMATION NOTICE 2011-20:

CONCRETE DEGRADATION BY ALKALI-SILICA

REACTION

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power reactor under

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of

Production and Utilization Facilities, except those who have permanently ceased operations

and have certified that fuel has been permanently removed from the reactor vessel.

All holders of or applicants for an early site permit, standard design certification, standard

design approval, manufacturing license, or combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

All holders of or applicants for a license for a fuel cycle facility issued pursuant to

10 CFR Part 70, Domestic Licensing of Special Nuclear Material.

All holders of and applicants for a gaseous diffusion plant certificate of compliance or an

approved compliance plan under 10 CFR Part 76, Certification of Gaseous Diffusion Plants.

All holders of and applicants for a specific source material license or for uranium recovery

operating license or construction permit under 10 CFR Part 40, Domestic Licensing of Source

Material. Uranium recovery facilities include conventional mills, heap leach facilities, and in situ

recovery facilities.

All holders of and applicants for an independent spent fuel storage installation license under

10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform

addressees of the occurrence of alkali-silica reaction (ASR)-induced concrete degradation of a

seismic Category 1 structure at Seabrook Station. The NRC expects that recipients will review

the information for applicability to their facilities and consider actions, as appropriate, to avoid

similar problems. However, suggestions contained in this IN are not NRC requirements;

therefore, no specific action or written response is required.

BACKGROUND

ASR is one type of alkali-aggregate reaction that can degrade concrete structures. ASR is a

slow chemical process in which alkalis, usually predominantly from the cement, react with

certain reactive types of silica (e.g., chert, quartzite, opal, and strained quartz crystals) in the

aggregate, when moisture is present. This reaction produces an alkali-silica gel that can absorb

water and expand to cause micro-cracking of the concrete. Excessive expansion of the gel can

lead to significant cracking which can change the mechanical properties of the concrete. In

order for ASR to occur, three conditions must be present: a sufficient amount of reactive silica

in the aggregate, adequate alkali content in the concrete, and sufficient moisture.

ASR can be identified as a likely cause of degradation during visual inspection by the unique

craze, map or patterned cracking and the presence of alkali-silica gel (see Figure 1 in the

enclosure). However, ASR-induced degradation can only be confirmed by optical microscopy

performed as part of petrographic examination of concrete core samples.

To prevent ASR-induced concrete degradation, the American Society for Testing and Materials

(ASTM) has issued standards for testing concrete aggregate during construction to verify that

only non-reactive aggregates are present. These standards include ASTM C227, Standard

Test Method for Potential Alkali Reactivity of Cement-Aggregate Combinations (Mortar-Bar

Method); ASTM C289, Standard Test Method for Potential Alkali-Silica Reactivity of

Aggregates (Chemical Method); ASTM C295, Standard Guide for Petrographic Examination of

Aggregates for Concrete; ASTM C1260, Standard Test Method for Potential Alkali Reactivity

of Aggregates (Mortar-Bar Method); ASTM C1293, Standard Test Method for Determination of

Length of Change of Concrete Due to Alkali-Silica Reaction; and ASTM C1567, Standard Test

Method for Determining the Potential Alkali-Silica Reactivity of Combinations of Cementitious

Materials and Aggregates (Accelerated Mortar-Bar Method).

ASR degrades the measured mechanical properties of the concrete at different rates.

Therefore, relationships between compressive strength and tensile or shear strength and

assumptions about modulus of elasticity that were used in the original design of affected

structures may no longer hold true if ASR-induced degradation is identified.

Technical information on ASR-induced concrete degradation appears in specialized literature, such as the U.S. Department of Transportation Federal Highway Administrations Report on the

Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction in Transportation Structures, issued January 2010, and the American Concrete Institutes ACI 221.1R-98, Report on Alkali

Reactivity.

DESCRIPTION OF CIRCUMSTANCES

After observing concrete cracking patterns typical of ASR, in August 2010, the licensee for

Seabrook Station performed petrographic examinations and compressive strength and modulus

of elasticity testing of concrete core samples removed from below-grade portions of the control

building (a seismic Category I structure) that confirmed that ASR had caused the cracking.

These concrete core samples demonstrated a substantial reduction in compressive strength compared to test cylinders cast during construction and a modulus of elasticity substantially

lower than the expected value. The licensee completed a prompt operability determination that

concluded margins to the code design limits remained such that the structural integrity of the

control building continued to be demonstrated.

The Seabrook Station final safety analysis report specifies concrete testing during construction

using ASTM C289 and ASTM C295, which were the accepted standards at the time of

construction. However, ASR-induced degradation still occurred.

The licensee believes that the waterproof membrane was damaged during original installation or

backfill activities causing water intrusion that resulted in the ASR problems. Water intrusion was

exacerbated by the fact that dewatering channels were abandoned.

Additional information appears in the licensees responses to requests for additional information

related to license renewal, dated December 17, 2010, April 14 and August 11, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML103540534, ML11108A131, and ML11227A023, respectively), and in NRC inspection reports

dated May 12 and May 23, 2011 (ADAMS Accession Nos. ML111330689 and ML111360432, respectively).

DISCUSSION

As noted above, ASTM has several standards for testing aggregates during construction to

verify that only non-reactive aggregates are present, thereby preventing future ASR-induced

degradation. However, ASTM issued updated standards ASTM C1260 and ASTM C1293 and

provided guidance in the appendices of ASTM C289 and ASTM C1293 that cautions that the

tests described in ASTM C227 and ASTM C289 may not accurately predict aggregate reactivity

when dealing with late- or slow-expanding aggregates containing strained quartz or

microcrystalline quartz. Therefore, licensees that tested using ASTM C227 and ASTM C289 could have concrete that is susceptible to ASR-induced degradation. Beginning at initial

construction, licensees may implement measures to prevent ASR-inducted concrete

degradation such as selecting non-reactive materials, and controlling water infiltration by

protecting and preserving waterproof membranes, or adding and maintaining dewatering

channels. Regardless of the measures taken during initial construction, visual inspections of

concrete can identify the unique map or patterned cracking and the presence of alkali-silica

gel in areas likely to experience ASR (i.e., concrete exposed to moisture). Additional

information can be found in the American Concrete Institutes ACI 349.3R-02, Evaluation of

Existing Nuclear Safety-Related Concrete Structures.

In 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear

Power Plants (the maintenance rule), the NRC requires that licensees monitor the performance

or condition of structures, systems, and components (SCCs) against licensee-established goals

in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling

their intended function. The regulations in 10 CFR 50.65 require that these goals be

established commensurate with safety and, where practical, take into account industry-wide

operating experience. In practice, for concrete structures, this usually translates into periodic

visual inspection; however, specific inspection criteria related to ASR are generally not included.

Section 1.5 of Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, explains that an acceptable structural monitoring program should evaluate the

results of periodic assessments to determine the extent and rate of any degradation of the

structures.

Once visual indications of ASR-induced concrete degradation have been identified, additional

actions to evaluate and monitor the condition, as recommended in the Federal Highway

Administration report (referenced above), may include confirming the presence of ASR through

microscopic examination of concrete cores; verifying the mechanical properties through testing

of concrete cores; and in situ monitoring of the concrete over time, such as crack mapping and

monitoring of concrete relative humidity. Nuclear power plant licensees may consider these

actions to determine the remaining potential reactivity, and the rate of ASR progression.

Because safety-related structures and nonsafety-related structures whose failure could affect

safety-related structures are within the scope of the maintenance rule, licensees are required to

monitor the condition of the structures against licensee-established goals to provide reasonable

assurance that the structures are capable of fulfilling their intended functions. If ASR-induced

degradation is identified in these structures, this condition monitoring would include determining

the extent and rate of the degradation.

The NRC staff is currently reviewing the license renewal application for Seabrook Station

submitted in accordance with 10 CFR 54, Requirements for Renewal of Operating Licenses for

Nuclear Power Plants. The Seabrook Station is the first plant to address ASR-induced

concrete degradation as part of license renewal. The licensee for Seabrook Station is

developing aging management programs that will include additional measures and actions to

manage the effects of aging from ASR-induced degradation during the period of extended

operation. In support of its license renewal application, the licensee for Seabrook Station will

submit additional information that the NRC staff will review to ensure the licensee develops an

acceptable program to manage the effects of ASR.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contact listed below or to the appropriate Office of Nuclear Reactor

Regulation project manager.

/RA by DWeaver for/

/RA/

Vonna Ordaz, Director

Timothy J. McGinty, Director

Division of Spent Fuel Storage

Division of Policy and Rulemaking

and Transportation

Office of Nuclear Reactor Regulation

Office of Nuclear Material Safety

and Safeguards

/RA by JTappert for/

Laura A. Dudes, Director

Division of Construction Inspection

and Operational Programs

Office of New Reactors

Technical Contact:

Bryce C. Lehman, NRR

301-415-1626 E-mail: Bryce.Lehman@nrc.gov

Enclosure:

Photograph of Concrete Degradation

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.

ML112241029 OFFICE

NRR/DLR/RASB

Tech Editor*

BC:NRR/DLR/RASB D: NRR/DLR

BC:NRO/DE/SEB1 NAME

BLehman

KAzariah-Kribbs

RAuluck

BHolian

BThomas

DATE

09/12/2011

09/29/2011 email

09/13/2011

09/22/2011

09/26/2011 email

OFFICE

BC: NRR/DE/EMCB LA: NRR/PGCB

PM:NRR/PGCB

BC:NRR/PGCB

NAME

MKhanna

CHawes

DBeaulieu

SRosenberg

DATE

09/12/2011 email

10/03/2011

09/29/2011

10/17/2011

OFFICE

D:NRO/DCIP

D:DSFST:NMSS

D:NRR/DPR

NAME

LDudes JTappert for V Ordaz

TMcGinty

OFFICE

10/21/2011

11/18/11

10/24/11

IN 2011-20 Photograph of Concrete Degradation

Figure 1 Patterned cracking indicative of ASR-induced degradation

(generic example-NOT from nuclear industry)