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. 
IN 2011-20 

Page 2 of 5 

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 Administration’s “Report on the 

Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction in Transportation Structures,” 

issued January 2010, and the American Concrete Institute’s 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 
IN 2011-20 

Page 3 of 5 

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 licensee’s 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 Institute’s 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 
IN 2011-20 

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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. 

 
IN 2011-20 

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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. 
IN 2011-20 

Page 5 of 5 

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. 

ADAMS Accession No.: 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 

 OFFICIAL RECORD COPY 
IN 2011-20 

Enclosure 

Page 1 of 1 

Photograph of Concrete Degradation 

Figure 1 Patterned cracking indicative of ASR-induced degradation 

(generic example–NOT from nuclear industry)