Improved Identification Techniques Against Alkali-Silica Reaction (ASR) Concrete Degradation at Nuclear Power Plants, 65023-65032 [2019-25489]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 84, Number 228 (Tuesday, November 26, 2019)]
[Proposed Rules]
[Pages 65023-65032]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-25489]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[Docket No. PRM-50-109; NRC-2014-0257]


Improved Identification Techniques Against Alkali-Silica Reaction 
(ASR) Concrete Degradation at Nuclear Power Plants

AGENCY: Nuclear Regulatory Commission.

ACTION: Petition for rulemaking; denial.

-----------------------------------------------------------------------

SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying a 
petition for rulemaking (PRM), PRM-50-109, dated September 25, 2014, 
submitted by the C-10 Research and Education Foundation (C-10 or the 
petitioner). The petitioner requests that the NRC amend its regulations 
to provide improved identification techniques for better protection 
against concrete degradation due to alkali-silica reaction (ASR) at 
U.S. nuclear power plants. The petitioner asserts that reliance on 
visual inspection will not adequately identify ASR, confirm ASR, or 
provide the current state of ASR damage without petrographic 
examination. The NRC is denying the petition because existing NRC 
regulations and NRC oversight activities provide reasonable assurance 
of adequate protection of public health and safety. Specifically, 
existing NRC regulations are sufficient to ensure that concrete 
degradation due to ASR will not result in unacceptable reductions in 
the structural capacity of safety-related structures at nuclear power 
plants.

DATES: The docket for the petition for rulemaking PRM-50-109 is closed 
on November 26, 2019.

ADDRESSES: Please refer to Docket ID NRC-2014-0257 when contacting the 
NRC about the availability of information regarding this petition. You 
can obtain publicly-available documents related to the petition using 
any of the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search on the petition Docket ID NRC-2014-0257. 
Address questions about NRC dockets to Carol Gallagher; telephone: 301-
415-3463; email: [email protected]. For technical questions, 
contact the individual listed in the FOR FURTHER

[[Page 65024]]

INFORMATION CONTACT section of this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
Supplementary Information section. For the convenience of the reader, 
instructions about obtaining materials referenced in this document are 
provided in Section V, Availability of Documents.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear 
Material Safety and Safeguards, telephone: 301-415-1519, email: 
[email protected], U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Table of Contents

I. The Petition
II. Public Comments on the Petition
III. Reasons for Denial
IV. Conclusion
V. Availability of Documents

I. The Petition

    On September 25, 2014, C-10, with assistance from the Union of 
Concerned Scientists (UCS), submitted a petition for rulemaking to the 
NRC (ADAMS Accession No. ML14281A124). The NRC docketed the petition on 
October 8, 2014, and assigned Docket No. PRM-50-109 to the petition. 
The petitioner requests that the NRC amend its applicable regulations 
to provide identification techniques for better protection against 
concrete degradation due to ASR at U.S. nuclear power plants. 
Specifically, the petitioner requests that the NRC require that all 
licensees comply with American Concrete Institute (ACI) Committee 
Report 349.3R, ``Evaluation of Existing Nuclear Safety-Related Concrete 
Structures'' (ACI 349.3R), and American Society for Testing and 
Materials (ASTM) Standard C856-11, ``Standard Practice for Petrographic 
Examination of Hardened Concrete'' (ASTM C856-11).
    The petitioner previously submitted a request for enforcement 
action in accordance with Sec.  2.206 of title 10 of the Code of 
Federal Regulations (10 CFR), ``Requests for action under this 
subpart,'' specific to Seabrook Station (ADAMS Accession No. 
ML16006A002). That petition was rejected by the NRC in a letter dated 
July 6, 2016 (ADAMS Accession No. ML16169A172), because the request 
addressed deficiencies within existing NRC rules, similar to those 
raised in PRM-50-109. While mention of Seabrook Station, which is the 
only nuclear power plant with a documented occurrence of ASR to date, 
is included in this document in response to the petitioner's comments, 
the NRC's focus in this denial is on the generic request that the NRC 
require that all licensees of nuclear plants comply with ACI 349.3R and 
ASTM C856-11.
    The petitioner raises the following three specific issues in PRM-
50-109.
    Issue 1: Visual inspections are not adequate to detect ASR, confirm 
ASR, or provide the current state of ASR damage.
    The petitioner asserts that visual inspections are not capable of 
adequately identifying ASR, confirming ASR, or providing accurate 
information on the state of ASR damage (i.e., its effect on structural 
capacity). The petitioner also asserts that only petrographic 
examinations (the use of microscopes to examine samples of rock or 
concrete to determine their mineralogical and chemical characteristics) 
in accordance with ASTM C856-11 are capable of determining or 
confirming whether ASR is present and determining the state of ASR 
damage. The petitioner offers additional information in five areas 
related to this issue.
    A. At an NRC public meeting at Seabrook Station on June 24, 2014, 
when C-10 asked if the NRC was investigating U.S. nuclear power plants 
for ASR concrete degradation, the NRC staff responded that ASR concrete 
degradation could be adequately identified through visual examination.
    B. When structural degradation is occurring, the petitioner asserts 
that it is critical to determine the root cause and confirm the form of 
degradation. The petitioner also asserts that the NRC has stated that 
ASR is confirmed only through petrographic examination, and in support 
of this statement the petitioner references an enclosure to a letter 
from the licensee for Seabrook Station, NextEra Energy Seabrook, LLC 
(NextEra) to the NRC, May 1, 2013 (ADAMS Accession No. ML13151A328).
    C. Commentaries by materials science expert Dr. Paul Brown, 
provided by C-10 and the UCS, challenge the central hypothesis in the 
report submitted by NextEra, ``Seabrook Station: Impact of Alkali-
Silica Reaction on Concrete Structures and Attachments'' (ADAMS 
Accession No. ML12151A397). As summarized in the petition, Dr. Brown 
challenges the conclusion in the report that ``confinement reduces 
cracking, and taking a core bore test would no longer represent the 
context of the structure once removed from the structure.''
    D. The petitioner also asserts that the NRC memorandum titled, 
``Position Paper: In Situ Monitoring of Alkali-Silica Reaction (ASR) 
Affected Concrete: A Study on Crack Indexing and Damage Rating Index to 
Assess the Severity of ASR and to Monitor ASR Progression'' (ADAMS 
Accession No. ML13108A047), supports the assertion that visual 
examination is insufficient to reliably identify ASR or evaluate its 
state (including contribution to rebar stress). The petitioner cites 
portions of the paper, which state that ASR can exist without 
indications of pattern cracking, visible surface cracking may be 
suppressed by heavy reinforcement while internal damage exists through 
the depth of the section, and crack mapping alone to determine ASR 
effects on the structure does not allow for the consideration of rebar 
stresses.
    E. Finally, the petitioner asserts that visual inspections are of 
limited scope and cannot identify areas of degradation in many portions 
of concrete structures, such as below-grade portions that cannot be 
visually examined but are most likely to be exposed to groundwater and 
be more vulnerable to ASR. The petitioner notes as an example cracking 
in the concrete wall of the shield building of the Davis-Besse Nuclear 
Power Station. This condition was discovered in 2011, when a hole was 
cut through the building's wall to replace the reactor vessel head, but 
had remained undetected by visual inspections for a long period.
    Issue 2: ACI and ASTM codes and standards address the detection and 
evaluation of ASR damage.
    The petitioner asserts that ACI 349.3R provides an acceptable means 
of protecting against excessive ASR concrete degradation and is 
endorsed by the NRC in Information Notice (IN) 2011-20, ``Concrete 
Degradation by Alkali-Silica Reaction'' (ADAMS Accession No. 
ML112241029). Quantitative criteria in ACI 349.3R can be used to 
evaluate inspection results. The petitioner also states that ASTM

[[Page 65025]]

C856-11 is an acceptable means of conducting petrographic examination.
    The petitioner also provided information specific to activities at 
Seabrook Station related to the implementation of ACI 349.3R and the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (BPV Code), Section XI, Subsection IWL. The petitioner 
states that ACI 349.3R requires the formation of a ``composite team,'' 
consisting of qualified civil or structural engineers, concrete 
inspectors, and technicians familiar with concrete degradation 
mechanisms and long-term performance issues, to effectively identify 
and evaluate concrete degradation, including degradation due to ASR.
    The petitioner claims that NextEra did not have a composite team as 
specified in ACI 349.3R, and since it became the owner of Seabrook 
Station, NextEra has not had a trained and dedicated ``responsible 
engineer'' conducting the inspections to accurately record the results 
or take further action as required. The petitioner asserts that NextEra 
failed to test the concrete despite the extent of cracking visibly 
increasing, and that NextEra never had a code-certified ``responsible 
engineer'' doing the visual inspections of the Seabrook containment in 
accordance with ASME BPV Code, Section XI, Subsection IWL.
    Issue 3: Regulations should require compliance with ACI 349.3R and 
ASTM C856-11.
    The petitioner states that, although both ACI 349.3R and ASTM C856-
11 are endorsed by the NRC, the NRC does not require nuclear power 
plant licensees to implement either of these standards.
    To support the position that use of the standards should be 
required, the petitioner offers Seabrook Station's ASR concrete 
degradation as an example that would have been identified before it 
caused moderate to severe degradation in seismic Category I structures 
if the NRC had required compliance with these existing standards. The 
petitioner claims that when NextEra determined 131 locations with 
``assumed'' ASR visual signs within multiple power-block structures 
during 2012, further engineering evaluations were not done. The 
petitioner also claims that, since discovering the situation, the NRC 
has not required Seabrook Station to: (1) Test a core bore taken from 
the containment; (2) use certified laboratory testing of key material 
properties to determine the extent of condition; or (3) obtain the data 
necessary to monitor the rate of progression.

II. Public Comments on the Petition

    The NRC published a notice of docketing of PRM-50-109 on January 
12, 2015 (80 FR 1476). The public comment period closed on March 30, 
2015. Comment submissions on this petition are available electronically 
via https://www.regulations.gov using docket number NRC-2014-0257.

Overview of Public Comments

    The NRC received 10 different comment submissions on the PRM. A 
comment submission is a communication or document submitted to the NRC 
by an individual or entity, with one or more individual comments 
addressing a subject or issue. Eight of the comment submissions were 
received during the public comment period. Two of the comment 
submissions were received after the comment period closed. The NRC 
determined that it was practical to consider the comment submissions 
received after the public comment period closed and considered all 10 
received. Key information for each comment submission is provided in 
the following table.

----------------------------------------------------------------------------------------------------------------
          Submission No.               ADAMS  accession No.           Commenter               Affiliation
----------------------------------------------------------------------------------------------------------------
1.................................  ML15026A339                 Josephine Donovan....  Private Citizen.
2.................................  ML15026A338                 Lynne Mason..........  Private Citizen.
3.................................  ML15027A178                 Katherine Mendez.....  Private Citizen.
4.................................  ML15076A457                 David Lochbaum.......  Union of Concerned
                                                                                        Scientists.
5.................................  ML15076A459                 Garry Morgan.........  Blue Ridge Environmental
                                                                                        Defense League--
                                                                                        Bellefonte Efficiency
                                                                                        and Sustainability Team/
                                                                                        Mothers Against
                                                                                        Tennessee River
                                                                                        Radiation (BREDL/BEST/
                                                                                        MATRR).
6.................................  ML15076A460                 G. Dudley Shepard....  Private Citizen.
7.................................  ML15085A523                 Jason Remer..........  Nuclear Energy Institute.
8.................................  ML15089A284                 James M. Petro, Jr...  NextEra Energy.
9.................................  ML15097A337                 Anonymous............  Anonymous.
10................................  ML15112A265                 Scott Bauer..........  STARS Alliance.
----------------------------------------------------------------------------------------------------------------

    Seven commenters expressed support for the PRM and proposed 
identification techniques, while the three remaining commenters 
(numbers 7, 8, and 10) opposed the PRM in part or in whole. Based on 
similarity of content, the public comments were grouped into six bins. 
The NRC reviewed and considered the comments in making its decision to 
deny the PRM. Summaries of each bin and the NRC's responses are 
provided in the following discussion in an order that provides 
appropriate context for the response to each of the comment bins.

NRC Responses to Comments on PRM-50-109

    Comment Bin 1: Existing inspection techniques will not adequately 
detect concrete degradation due to ASR, and C-10's proposed solutions 
(i.e., requiring compliance with ACI 349.3R and ASTM C856-11 via 
regulation) are appropriate to adequately detect ASR degradation. 
(Submission 4, Submission 5, Submission 6)
    NRC Response: Although the NRC agrees with the petitioner that 
visual inspections are not enough to positively confirm ASR, the staff 
finds visual inspection sufficient to detect ASR concrete degradation 
before the safety function of a structure or component would be 
significantly degraded. The NRC disagrees with the comments that ACI 
349.3R and ASTM C856-11 should be regulatory requirements. The current 
ASR literature and case history, as described in Section III and 
referenced in Section V, ``Availability of Documents,'' of this 
document, provide no evidence that ASR would degrade the safety 
function of a structure or component before it expands to a degree that 
would cause visible symptoms, such as cracking. Existing regulations 
require inspection methods that can detect applicable degradation 
mechanisms (including ASR) and require that significant degradation 
regardless of cause be addressed appropriately through additional 
plant-specific inspections or structural evaluations. Furthermore, the 
documents (ACI 349.3R and ASTM C856-11) do not provide specific 
guidance for identifying ASR

[[Page 65026]]

degradation in structures. Therefore, requiring their use via 
regulation would not provide improved techniques for identifying ASR 
degradation. Additional details on the NRC's position can be found in 
Section III, ``Reasons for Denial,'' of this document.
    Comment Bin 2: The NRC should grant the C-10 petition for 
rulemaking because visual inspection of ASR concrete degradation is 
insufficient. (Submission 1, Submission 2)
    NRC Response: The NRC disagrees with this comment. As indicated in 
the response to Comment Bin 1, there is no evidence in current ASR 
literature and case history that ASR would degrade the safety function 
of a structure or component before it expands to a degree that would 
cause visible symptoms. In addition, NRC staff finds visual inspection 
sufficient to detect ASR concrete degradation before the safety 
function of a structure or component would be degraded. Moreover, the 
commenters did not provide a basis for their position that visual 
inspection of concrete degradation is insufficient to identify ASR that 
would lead to unacceptable changes in concrete structural properties.
    Comment Bin 3: The NRC should investigate the concrete cracks at 
Seabrook Station because the concrete degradation poses serious safety 
concerns. (Submission 3)
    NRC Response: The NRC views this comment as a request for 
regulatory action outside the scope of PRM-50-109. As discussed in 
Section III of this document, the NRC has referred this comment to its 
Region I allegations staff, and has advised the commenter of this 
request.
    Comment Bin 4: The nuclear industry does not believe that 
rulemaking is necessary to resolve issues related to inspecting 
concrete for ASR degradation. Following the issuance of NRC IN 2011-20, 
licensees took appropriate actions by: (a) Recording the issue in the 
Institute for Nuclear Power Operations Operating Experience system; and 
(b) updating their Structures Monitoring Program, improving procedures, 
and informing responsible individuals concerning examination for 
conditions that could potentially indicate the presence of ASR. In 
addition, there already exist ample regulatory requirements to ensure 
appropriate attention is given to potentially degraded concrete, 
including due to ASR. (Submission 7, Submission 10)
    NRC Response: The NRC agrees with the comment. By issuing IN 2011-
20, the NRC made the U.S. nuclear power industry aware of the operating 
experience related to ASR concrete degradation at Seabrook Station. 
Licensees are expected to evaluate INs in their operating experience 
programs and to incorporate, as appropriate and applicable, the 
information into their monitoring programs and procedures. Multiple 
license renewal applications (LRAs) submitted after the issuance of IN 
2011-20 included information that demonstrates the monitoring programs 
have been updated to inspect for ASR degradation, regardless of the 
aggregate reactivity test results from construction (see, for example, 
Section 3.5.2.2.2.1.2 of LaSalle County Station LRA (ADAMS Accession 
No. ML14343A849), Waterford Steam Electric Station LRA (ADAMS Accession 
No. ML16088A324), and River Bend Station LRA (ADAMS Accession No. 
ML17153A282)).
    Existing regulations such as Sec.  50.55a, ``Codes and Standards''; 
Sec.  50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants''; 10 CFR part 50, appendix B, 
``Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants''; 10 CFR part 50, appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors''; and 10 
CFR part 54, ``Requirements for Renewal of Operating Licenses for 
Nuclear Power Plants,'' require licensees to monitor the performance or 
condition of structures and take corrective action to address degraded 
or nonconforming conditions in a manner commensurate with the safety 
significance of the structures. Compliance with these regulations 
provides reasonable assurance that affected structures remain capable 
of performing their intended functions. Further, the NRC confirms the 
acceptability of licensees' approaches through processes such as the 
reactor oversight process, license renewal, and review of licensees' 
responses to generic communications (e.g., bulletins, generic letters, 
and INs that address significant industry events, operating experience, 
and degradation-specific issues that may have generic applicability). 
The existing regulatory requirements and processes provide reasonable 
assurance of adequate protection of public health and safety against 
the potential results of degradation of concrete structures; therefore, 
it is not necessary to amend the NRC's regulations.
    The technical comments and clarifications made by the commenters 
related to ACI 349.3R and the role of visual inspections are addressed 
in Section III of this document.
    Comment Bin 5: New rulemaking is not necessary to resolve issues 
related to inspecting concrete for ASR. The ACI 349.3R and ASTM C856-11 
have been used for investigation of ASR conditions at Seabrook Station; 
however, neither standard provides inspectors with new or improved 
means to identify, monitor, or assess ASR-impacted structures, as 
implied by the petition. The commenter questions the basis of the 
petition, including misconceptions and factual errors made in the 
petition concerning NextEra activities at Seabrook Station. (Submission 
8)
    NRC Response: The NRC agrees with the comment that new rulemaking 
is not needed. The guidance in ACI 349.3R is primarily based on visual 
inspection, addresses only commonly occurring degradation conditions in 
nuclear structures, and provides very limited guidance with regard to 
ASR identification, monitoring, and evaluation. Therefore, it is not 
considered an authoritative document for ASR. ASTM C856-11 is a 
consensus standard that provides an established method for conducting 
petrography that can be used to confirm the diagnosis of ASR. Neither 
ACI 349.3R nor ASTM C856-11, however, provides a method for monitoring 
progression, or evaluating and quantifying observed ASR effects on 
structural capacity or performance. These documents have been in 
existence since 1996 (for ACI 349.3R) and 1977 (for ASTM C856-11) and 
do not provide any new or improved methods beyond what is already 
standard practice in the concrete industry.
    The portions of the comment concerning NextEra activities at 
Seabrook Station are addressed in Section III of this document.
    Comment Bin 6: Current ASME testing protocols should be followed. 
Ultrasonic testing should be conducted for reactor pressure vessels to 
test for defects and radiation filters should be installed on pressure 
vents as a post-Fukushima precaution. (Submission 9)
    NRC Response: As stated in Section III of this document, Section 
50.55a(g)(4) requires compliance with the ASME BPV Code, Section XI. 
The ASME BPV Code, Section XI, Subsection IWL, provides techniques for 
examination and evaluation of concrete surfaces that licensees follow 
under their licensing bases. The comments pertaining to ultrasonic 
testing of reactor pressure vessels and installation of radiation 
filters are not related to ASR degradation and are outside the scope of 
PRM-50-109.

III. Reasons for Denial

    The NRC has determined that rulemaking, as requested in the 
petition, is not needed for reasonable assurance

[[Page 65027]]

of adequate protection of public health and safety at nuclear power 
plants with respect to ASR. The NRC's evaluation of the three issues 
raised in PRM-50-109 are set forth below.
    Issue 1: Visual Inspections are not adequate to detect ASR, confirm 
ASR, or provide the current state of ASR damage.
    The NRC agrees with the petitioner that visual inspections are not 
enough to positively confirm ASR. However, given the slow progression 
of ASR, visual inspections are sufficient to identify manifestations of 
potentially damaging ASR before the safety function of a structure or 
component would be degraded. This would be sufficient to inform whether 
further actions should be taken. Therefore, the NRC's position is that 
visual examination is acceptable for routinely monitoring concrete 
structures to identify areas of potential structural distress or 
degradation, including degradation due to ASR. This position is 
supported by the current ASR literature and case history, as referenced 
in Section V of this document. The occurrence of ASR expansion results 
in one or more common visual indications (e.g., expansion causing 
deformation, movement, or displacement; cracking; surface staining; gel 
exudations; pop-outs) prior to causing significant structural 
degradation (as shown in Federal Highway Administration (FHWA)-HIF-09-
004 and Canadian Standards Association (CSA) A864-00, referenced in 
Section V of this document). However, the presence of one or more of 
these visual symptoms is not necessarily an indication that ASR is the 
main factor responsible for the observed symptoms. If there are visual 
indications, the presence or absence of ASR should be confirmed by an 
acceptable method such as petrographic examination.
    Based on this information, the NRC maintains that visual 
examination is an acceptable method for detecting indications of ASR 
degradation. Once ASR is suspected based on visual indications, the 
licensee would need to conduct additional inspections, testing (non-
destructive or invasive), petrographic analysis, or structural 
evaluations, as appropriate to the specific case, to evaluate the 
effects of ASR on structural performance under design loads. This 
general approach is similar to and consistent with the approach 
recommended in literature related to ASR (e.g., FHWA-HIF-09-004 and 
guidance by the Institution of Structural Engineers, referenced in 
Section V of this document).
    The NRC evaluated the following five areas in which the petitioner 
provided additional information related to this issue.
    A. Regarding the statements made by the NRC staff during the June 
24, 2014, public meeting the NRC staff stated that it finds the use of 
visual examination acceptable for routine periodic monitoring, in 
implementing a structures monitoring program under Sec.  50.65 and the 
containment inservice inspection program under Sec.  50.55a, and in 
identifying the general condition of concrete structures and areas that 
are suspected to have deterioration or distress due to any degradation 
mechanism, including ASR. If the licensee identifies visual indications 
of ASR, the next step would be to confirm ASR by petrographic 
examination or other acceptable methods, and conduct further 
assessments, as necessary, to determine the impact on the structure's 
intended functions and the need for corrective actions, as required by 
appendix B to 10 CFR part 50. While visual inspections alone would not 
confirm the presence or absence of ASR, a petrographic examination of 
concrete is not necessary prior to manifestation of visual symptoms of 
ASR, given the minimal impact ASR has on structural performance of 
reinforced concrete structures at this stage. The NRC maintains its 
position that visual examination is an acceptable approach for 
assessing the concrete's general condition and identifying areas of 
potential structural distress or deterioration, including areas where 
ASR is suspected.
    B. Specific to the petitioner's statement related to the need to 
determine the root cause of degradation, existing NRC regulations 
require that licensees promptly identify conditions adverse to quality, 
determine the cause, and take corrective actions. Specifically, 
Criterion XVI, ``Corrective Action,'' of 10 CFR part 50, appendix B 
requires that conditions adverse to quality such as failures, 
malfunctions, deficiencies, deviations, defective material and 
equipment, and nonconformances are promptly identified and corrected. 
In the case of significant conditions adverse to quality, the measures 
shall assure that the cause of the condition is determined and 
corrective action taken to preclude repetition. The NRC agrees that, 
while other techniques may emerge, petrographic examination of the 
concrete sample under a microscope is a well-established technique to 
confirm the presence or absence of ASR at any stage.
    Once ASR is confirmed at a site by petrographic examination 
(conducted after manifestation of characteristic visual symptoms), it 
is conservative to assume that other structures exhibiting visible 
symptoms are also affected, based on similarity of materials and 
environmental exposure conditions. The degradation can then be 
addressed accordingly.
    Appendix B to 10 CFR part 50 already requires the identification of 
a significant condition adverse to quality, the determination of the 
cause of the condition through root cause analyses and appropriate 
follow-up corrective actions. Therefore, a generic revision to the 
NRC's regulations is not necessary.
    C. The NRC has previously responded to the statements referenced by 
the petitioner from Dr. Paul Brown, which were included in a letter 
from UCS to the NRC dated November 4, 2013 (ADAMS Accession No. 
ML13309B606). In a December 6, 2013 response (ADAMS Accession No. 
ML13340A405), the NRC noted that information from drilled cores may be 
valuable for assessing the impact of ASR on concrete; however, the use 
of test data from cores alone may not be an appropriate, realistic 
indicator of overall structural performance.
    Additionally, the NRC notes that ASR literature and case history 
indicate that ASR has a much more detrimental effect on the mechanical 
properties of concrete cores and cylinders than on the structural 
behavior of reinforced concrete structural components and systems (as 
described in TXDOT Technical Report No. 12-8XXIA006 and the ACI 
Structural Journal article referenced in Section V of this document). 
These documents indicate that the empirical relationships in the ACI 
codes between concrete-cylinder compressive strength and other 
mechanical properties, including structural capacity, may not 
necessarily remain valid for ASR-affected structures. Reinforced 
concrete structures and components respond to load as part of a 
composite structural system in which there are external restraints, 
internal confinement, and interaction between the steel reinforcement 
and the concrete. Therefore, an evaluation of the impact of ASR on 
performance of affected reinforced concrete structural components and 
systems should consider the context to obtain a realistic assessment of 
the impact on structural capacity. The use of core test data in the 
traditional manner, alone, may not be appropriate or realistic to 
assess structural performance of ASR-affected structures.
    D. Regarding the petitioner's reference to the NRC position paper 
(ADAMS

[[Page 65028]]

Accession No. ML13108A047), that document is not an official NRC 
position on the topic, but rather was prepared by an individual staff 
member to facilitate internal technical discussion and inform staff 
review of an issue. The NRC's current position on the role of visual 
inspections in identifying ASR is set forth in this document. The 
referenced position paper does not state that visual examination is 
insufficient to identify indications of ASR. However, it does note that 
surface cracking or crack mapping, alone, may not indicate the severity 
of ASR degradation and is not adequate to determine structural effects 
of ASR. The NRC agrees that surface crack mapping alone is not adequate 
to monitor ASR progression and to address its structural effects. In 
addition, petrographic examination provides very limited information to 
evaluate the structural effects of ASR.
    Addressing visual indications of a potential concrete-degradation 
issue does not end with the visual inspection. Under existing NRC 
regulations, if indications of distress or deterioration are visually 
identified, licensees are required to address the effects of the 
observed degradation and demonstrate that the structure remains capable 
of performing its safety functions. Depending on the observed 
conditions, this can be accomplished through additional inspections, 
testing, structural evaluations, or a combination thereof.
    E. Specific to the petitioner's comment on the limited scope of 
visual inspections, the NRC agrees that visual inspections cannot 
directly identify degradation in inaccessible portions of concrete 
structures. However, many below-grade structures in nuclear power 
plants are accessible for visual inspection on the interior face of the 
concrete. Additionally, ASR degradation or expansion in inaccessible 
areas would manifest visually in accessible areas, in the form of 
cracking, displacements, or deformations, before causing a significant 
structural impact. As noted previously, current ASR literature and case 
history show that visual inspections are sufficient to identify 
manifestations of potentially damaging ASR before there would be 
significant structural impacts. For concrete containment structures, 
existing regulations in Sec.  50.55a(b)(2)(viii) require evaluation of 
the acceptability of inaccessible areas when conditions exist in 
accessible areas that could indicate the presence of, or could result 
in, degradation to such inaccessible areas. Therefore, existing 
regulations, regulatory guidance, and licensee programs have provisions 
to adequately address degradation in inaccessible areas.
    The issue of laminar cracking in the shield building at Davis-
Besse, referenced by the petitioner, has no connection to ASR 
detection. Davis-Besse was a unique situation resulting from a 
combination of extreme environmental conditions and the design 
configuration of the shield building. The licensee evaluated the issue, 
including operability determinations and root cause analysis in its 
corrective action program; and the NRC's continued oversight of the 
issue has been documented in a series of NRC inspection reports, the 
latest of which is IR 05000346/2014008, dated May 28, 2015 (ADAMS 
Accession No. ML15148A489).
    Issue 2: Codes and standards exist for detecting and evaluating ASR 
damage.
    The NRC disagrees that there are consensus codes or standards 
sufficient to provide guidance for detecting and evaluating ASR damage. 
The scope of both ACI 349.3R and ASTM C856-11 are discussed separately 
below.
    A. The ACI 349.3R is an ACI committee technical report intended to 
provide recommended guidance for developing and implementing a 
procedure for inspection and evaluation of many common concrete 
degradation mechanisms in nuclear concrete structures. It contains only 
very limited general information regarding ASR. ASR is not a common 
condition in nuclear power plants, and the quantitative evaluation 
criteria provided in the document have little or no specific 
applicability to ASR degradation. Therefore, ACI 349.3R is not an 
authoritative document to address and evaluate the impact of ASR on 
intended functions of affected structures.
    The discussion of evaluation techniques in ACI 349.3R recommends 
visual inspection as the initial technique used for any evaluation, and 
states that visual inspection can provide significant quantitative and 
qualitative data regarding structural performance and the extent of any 
degradation. The recommended approach places emphasis on the use of 
general condition survey practices (visual inspection) in the 
evaluation, supplemented by additional testing or analysis as needed, 
based on the results of the general survey. Chapter 5, ``Evaluation 
Criteria,'' of ACI 349.3R states: ``these guidelines focus on common 
conditions that have a higher probability of occurrence and are not 
meant to be all-inclusive. These criteria primarily address the 
classification and treatment of visual inspection findings because this 
technique will have the greatest usage.''
    Although ACI 349.3R provides useful general guidance for the 
development and implementation of a monitoring plan for concrete 
structures, the NRC has neither formally endorsed nor approved it for 
use. Instead, IN 2011-20 simply mentions ACI 349.3R as a resource where 
additional information may be found regarding visual inspections (ADAMS 
Accession No. ML112241029). Since ASR degradation would need to be 
addressed on a degradation-specific and plant-specific basis, requiring 
the use of ACI 349.3R would not provide better protection against ASR 
concrete degradation than the current NRC requirements.
    Related to the petitioner's comments on ``composite teams,'' the 
NRC agrees that qualified personnel should be used to conduct 
activities pertaining to safety-related functions of structures, 
systems, and components (SSCs). Existing regulations provide for this 
in the quality assurance program requirements under appendix B to 10 
CFR part 50. This appendix requires applicants and licensees to 
establish and implement a quality assurance program that applies to all 
activities affecting the safety-related functions of SSCs. This program 
specifies controls to provide adequate confidence that SSCs will 
perform satisfactorily in service, including appropriate qualification 
and training of personnel performing activities affecting quality to 
assure suitable proficiency. This adequate confidence is part of the 
basis for concluding that reasonable assurance of adequate protection 
is provided. The ASME BPV Code, Section XI, Subsection IWL, defines 
specific qualifications and responsibilities of the ``responsible 
engineer,'' who evaluates the examination results and the condition of 
the structural concrete related to the containment. Section 
50.55a(g)(4) requires compliance with the ASME BPV Code, Section XI. In 
addition to Sec.  50.55a requirements for containments, safety-related 
structures are monitored under Sec.  50.65 (the maintenance rule), and 
the associated qualification requirements are typically provided in the 
licensee's implementing procedures, based on their 10 CFR part 50, 
appendix B program.
    As for the petitioner's claim related to the implementation of ACI 
349.3R at Seabrook Station, including the formation of a composite 
team, this topic is outside the scope of the NRC's consideration of the 
generic rulemaking action in response to PRM-50-109. However, this 
apparent claim of licensee wrongdoing was considered by

[[Page 65029]]

the NRC's allegations staff in Region I. After discussions with the 
petitioner, it was confirmed that the petitioner cited the issues with 
NextEra as examples of its concerns with regulations and did not intend 
the issues to be considered as allegations.
    B. Regarding the petitioner's comments on ASTM C856-11, although 
the NRC has neither formally endorsed nor approved its use, the NRC 
agrees that ASTM C856-11 is a consensus standard that details how to 
conduct petrographic analysis of concrete bores and provides an 
acceptable method to positively confirm the diagnosis of ASR. However, 
it does not provide any guidance on when cores should be taken, from 
where cores should be taken, how many cores should be taken, or how 
frequently cores should be taken. Also, it does not provide a method to 
evaluate ASR damage for impact on structural performance.
    ASTM C856-11 outlines procedures for the petrographic examination 
of samples of hardened concrete for a variety of purposes. One of the 
purposes of this consensus standard is identifying visual evidence to 
establish whether ASR has taken place, what aggregate constituents were 
affected, and what evidence of the reaction exists. Petrographic 
examination provides an assessment of the extent of ASR gel development 
and its intrusion into the pores of the concrete sample; however, 
petrographic examination does not indicate the impact of the ASR 
reaction on the structural performance under design loads. Furthermore, 
ASTM C856-11 does not provide any guidance on monitoring or evaluating 
a concrete structure, such as when to take cores, or which portion of a 
structure should be evaluated via core bores.
    Materials laboratories that perform petrographic examination of 
hardened concrete samples typically follow the current ASTM C856 
standard practice for the application, unless another specific 
procedure is specified in the request. The standard to which a plant-
specific petrographic examination is performed is specified by the 
licensee and not addressed in the regulations. However, appendix B to 
10 CFR part 50 requires licensees to ensure that activities affecting 
safety-related functions are controlled to provide adequate confidence 
that SSCs will perform satisfactorily in service. Also, 10 CFR part 50, 
appendix A, ``General Design Criteria for Nuclear Power Plants,'' 
Criterion 1, ``Quality standards and records,'' requires, in part, that 
``where generally recognized codes and standards are used, they shall 
be identified and evaluated to determine their applicability, adequacy, 
and sufficiency and shall be supplemented or modified as necessary to 
assure a quality product in keeping with the required safety 
function.'' Therefore, the licensee must ensure the analysis is 
sufficient to identify ASR.
    In summary, both ACI 349.3R and ASTM C856-11 provide useful 
guidance and methods licensees may adopt, as applicable, to meet 
requirements in existing NRC regulations, such as Sec.  50.55a, Sec.  
50.65, and 10 CFR part 54. However, neither of the documents provide 
methods to comprehensively address the long-term structural impact and 
management of ASR degradation.
    Issue 3: Regulations should require compliance with ACI 349.3R and 
ASTM C856-11.
    The NRC disagrees that its regulations need to be revised to 
require compliance with ACI 349.3R and ASTM C856-11. The NRC's existing 
regulations are sufficient to provide reasonable assurance of adequate 
protection of public health and safety due to concrete degradation, 
including ASR.
    The petition does not take into account the NRC's existing 
regulatory requirements that each nuclear power reactor licensee must 
meet to demonstrate the ongoing capability of structures to perform 
their intended safety functions. The NRC's regulatory requirements are 
applicable to all operating reactors and focused on overall structure 
and component performance requirements necessary to maintain intended 
safety functions. The NRC's regulations do not typically prescribe how 
licensees must meet the requirements, nor do the regulations normally 
address degradation-specific issues. The following discussion 
identifies and briefly summarizes the relevant regulatory requirements 
and processes and explains how they require licensees to address ASR 
before it becomes a safety issue.
     Section 50.65 requires licensees to monitor the 
performance or condition of SSCs under its scope, including safety-
related structures, considering industry-wide operating experience, in 
a manner sufficient to provide reasonable assurance that these SSCs are 
capable of fulfilling their intended functions. For structures, this 
requirement is normally met by periodically monitoring their condition 
on a frequency that is commensurate with their safety significance and 
condition. If the basic assessments identify degradation, additional 
degradation-specific condition monitoring is required, along with more 
frequent assessments until the degradation is addressed. Regulatory 
Guide (RG) 1.160, ``Monitoring the Effectiveness of Maintenance at 
Nuclear Power Plants,'' provides guidance on methods acceptable to the 
NRC staff for implementation of the maintenance rule and includes the 
attributes of an acceptable structural monitoring program. In summary, 
Sec.  50.65 already requires structural assessments that are adequate 
to detect visual indications of ASR before it would pose a significant 
structural concern.
     Criterion XVI, ``Corrective Action,'' of appendix B to 10 
CFR part 50 requires licensees to implement a corrective action program 
to assure that conditions adverse to quality and non-conformances are 
promptly identified and corrected. In the case of significant 
conditions adverse to quality, the measures shall assure that the cause 
of the condition is determined, and corrective action is taken to 
preclude repetition. This requirement applies to all degradation 
mechanisms, including ASR. In the case of ASR, a licensee would have to 
identify the root cause of the degradation and address the degradation, 
such that intended safety functions are not impacted. Accordingly, 
Criterion XVI is an NRC regulatory requirement that provides for the 
identification and further technical evaluation of ASR, before there 
would be significant degradation to the structural integrity of safety-
related concrete structures at nuclear power plants.
     Section 50.55a(g)(4) requires licensees to inspect 
concrete containments in accordance with the ASME BPV Code, Section XI, 
Subsection IWL, as incorporated by reference and subject to conditions. 
Subsection IWL requires that a general visual examination of all 
accessible containment concrete surfaces be conducted every 5 years by 
qualified personnel under the direction of the ``responsible 
engineer.'' Further, Subsection IWL requires a detailed visual 
examination to determine the magnitude and extent of deterioration and 
distress of suspect containment concrete surfaces initially detected by 
general visual examinations. Subsection IWL specifies acceptance 
standards based on acceptance by examination, acceptance by engineering 
evaluation (requires preparation of an engineering evaluation report 
including cause of the condition), or acceptance by repair/replacement. 
In accordance with the condition on use of Section XI in Sec.  
50.55a(b)(2)(viii)(E), licensees must evaluate the acceptability of 
inaccessible areas when conditions exist in accessible areas that could 
indicate the presence of or result in degradation to such inaccessible 
areas. These

[[Page 65030]]

requirements are designed to ensure that visual indications of ASR will 
be detected prior to causing significant structural degradation that 
could impact the intended safety function of the containment. 
Accordingly, Sec.  50.55a is a requirement that provides for the 
identification and further technical evaluation of ASR, before there 
would be significant degradation of structural integrity of concrete 
containment structures at nuclear power plants.
     Appendix J to 10 CFR part 50, ``Primary Reactor 
Containment Leakage Testing Requirements for Water Cooled Reactors,'' 
requires that primary reactor containments periodically meet the 
leakage-rate test requirements to ensure that (a) leakage does not 
exceed allowable rates listed in the technical specifications; and (b) 
integrity of the containment structure is maintained during its service 
life. This regulation requires periodic performance monitoring of the 
containment to demonstrate that the containment can perform its 
intended safety function, regardless of identified degradation. If the 
containment were unable to meet the requirements of 10 CFR part 50, 
appendix J, it would be declared inoperable and the plant could not 
return to operation until the issue was addressed. Accordingly, 
appendix J of 10 CFR part 50 is a regulatory requirement that provides 
for the identification and technical evaluation of ASR, before there 
would be significant degradation of structural integrity of concrete 
containment structures at nuclear power plants.
     Section 54.21(a)(3) requires applicants for license 
renewal to demonstrate that the effects of aging will be adequately 
managed, such that the intended functions of structures and components 
subject to aging management are maintained, consistent with the current 
licensing basis for the period of extended operation. Regulatory 
guidance for developing aging management programs, including for ASR 
aging effects on concrete structures, is provided in NUREG-1801, 
``Generic Aging Lessons Learned Report'' (GALL Report). Any licensee 
applying for license renewal must have a structural aging management 
program in place that can identify indications of concrete degradation, 
including degradation due to ASR, before it becomes an issue that could 
impact an intended safety function. Accordingly, Sec.  54.21(a)(3) is a 
regulatory requirement that provides for the identification and further 
technical evaluation of ASR, before there is significant degradation to 
the structural integrity of safety-related concrete structures at 
nuclear power plants.
     The Reactor Oversight Process (ROP) is the process that 
the NRC uses to verify that power reactors are operating in accordance 
with NRC rules and regulations. Under the ROP, the NRC conducts routine 
baseline inspections, problem identification and resolution 
inspections, reactive inspections, and other assessments of plant 
performance. If licensees are not properly meeting the regulations, the 
NRC can take actions to protect public health and safety.
     The generic communications process is used to address 
potential generic issues that are safety significant and may 
necessitate action by licensees to resolve. Generic communications, 
which include bulletins, generic letters and INs, are used to convey 
safety significant issues and operating experience, including 
degradation-specific issues. The NRC has issued a generic communication 
(IN 2011-20) to inform the industry of the generic impacts of ASR. 
Information about the NRC's Generic Communications Program is available 
at https://www.nrc.gov/about-nrc/regulatory/gencomms.html.
     The enforcement process may be used if licensees fail to 
adequately address safety-significant issues, consistent with the 
regulatory requirements as outlined above. The NRC may use enforcement 
actions, including issuing orders pursuant to Sec.  2.202, ``Orders,'' 
to modify, suspend, or revoke a license if ASR becomes a safety-
significant issue that a licensee is not adequately addressing.
    In addition to these generic requirements and processes, the GALL 
Report (NUREG-1801) makes specific reference to ACI 349.3R in its 
guidance for aging management programs (AMPs). AMP XI.S6, ``Structures 
Monitoring,'' recommends that visual inspection be used to identify 
structural distress or deterioration of concrete, such as that 
described in ACI 201.1R and ACI 349.3R. In addition, the GALL Report 
notes that the personnel qualifications in Chapter 7 and the evaluation 
criteria in Chapter 5 of ACI 349.3R are acceptable for concrete 
structures. However, the GALL Report also notes that use of plant-
specific criteria may also be justified. Although ACI 349.3R is one 
acceptable method to monitor concrete structures for degradation, it is 
not the only method, and so there is no need for the NRC to require its 
exclusive use via regulation.
    With respect to ASTM C856-11, the NRC agrees that it is an 
acceptable and established consensus testing standard for conducting 
petrographic examination of hardened concrete that can be used to 
confirm the diagnosis of ASR. However, as discussed previously, the 
NRC's existing regulations in 10 CFR part 50, appendix A and appendix 
B, ensure appropriate methods or standards are used when conducting 
tests associated with safety-related structures. Therefore, there is no 
need to require the use of ASTM C856-11 through regulation.
    The NRC also considered whether ASR concrete degradation raises new 
safety concerns that would justify additional regulatory requirements 
for all licensees beyond those already included in NRC regulations. 
While it is possible that there could be plants that used a potentially 
reactive aggregate in their concrete, the NRC is not aware of any U.S. 
nuclear power plants, other than Seabrook Station, that have a 
documented occurrence of ASR. The NRC notes that the use of a 
potentially reactive aggregate does not necessarily result in the 
occurrence of ASR. In addition to reactive aggregates, relatively high 
alkali content in the cement, and high relative humidity levels are 
necessary for ASR to occur. Through the issuance of IN 2011-20, the NRC 
has informed licensees of the occurrence of ASR-induced concrete 
degradation at Seabrook Station, with the expectation that the 
operating experience would be evaluated by licensees and considered for 
appropriate action. Thus, the nuclear power industry is aware of the 
potential for ASR to occur, even if aggregates were screened out based 
on reactivity or other tests conducted at the time of construction. For 
the reasons outlined above, the NRC has determined that the agency's 
existing regulatory structure is sufficient for the identification and 
technical evaluation of ASR before there is significant degradation to 
the structural integrity of safety-related concrete structures at 
nuclear power plants. Therefore, new or amended regulations are not 
needed to require industry-wide compliance with ACI 349.3R and ASTM 
C856-11.
    The petitioner's claims related to Seabrook Station are outside the 
scope of the NRC's consideration of the generic rulemaking action in 
response to PRM-50-109; however, the apparent claims of NRC wrongdoing 
were forwarded to the NRC's Office of the Inspector General and 
subsequently to the NRC's allegations staff in Region I. After 
discussions with the petitioner, the NRC confirmed that the petitioner 
cited the issues as examples of their concerns with the regulations and 
did

[[Page 65031]]

not intend them to be considered as allegations or claims of 
wrongdoing.

IV. Conclusion

    For the reasons cited in Section III of this document, the NRC is 
denying PRM-50-109 under Sec.  2.803. Existing NRC regulations 
establish programmatic and design basis requirements that are adequate 
to address the effects of concrete degradation mechanisms, including 
ASR, in safety-related structures. Compliance with these regulations, 
verified through NRC licensing and oversight processes, provide 
reasonable assurance of adequate protection of public health and 
safety. Specifically, existing NRC regulations ensure that concrete 
degradation due to ASR will not result in unacceptable reductions in 
structural capacity of safety-related structures at nuclear power 
plants. Therefore, new or amended regulations to require the use of the 
documents identified in the PRM (ACI 349.3R and ASTM C856-11) to 
provide better protection against concrete degradation due to ASR are 
not needed in order to provide reasonable assurance of adequate 
protection of public health and safety at U.S. nuclear power plants.

V. Availability of Documents

    The documents identified in the following table are available to 
interested persons through one or more of the following methods, as 
indicated. For more information on accessing ADAMS, see the ADDRESSES 
section of this document.

------------------------------------------------------------------------
                              ADAMS Accession No./
                                Federal Register
          Document             citation/report No.   Link to publication
                                    and date
------------------------------------------------------------------------
                              PRM Documents
------------------------------------------------------------------------
PRM from the C-10 Research    ADAMS Accession No.   https://
 and Education Foundation.     ML14281A124,          pbadupws.nrc.gov/
                               September 25, 2014.   docs/ML1428/
                                                     ML14281A124.pdf.
Federal Register notice for   Federal Register/     https://www.gpo.gov/
 PRM, notice of docketing,     Vol. 80, No. 7/       fdsys/pkg/FR-2015-
 and request for comment.      Monday, January 12,   01-12/html/2015-
                               2015/Proposed Rules.  00199.htm.
SECY-18-0036, ``Denial of     ADAMS Accession No.   https://
 Petition for Rulemaking       ML15301A084, March    pbadupws.nrc.gov/
 Submitted by the C-10         8, 2018.              docs/ML1530/
 Foundation (PRM-50-109).                            ML15301A084.pdf.
------------------------------------------------------------------------
Public Comments on PRM (see table under the heading, I. Public Comments
 on the Petition).
------------------------------------------------------------------------
                     ASR-Related Technical Materials
------------------------------------------------------------------------
``Standard Practice for       ASTM C856-11, 2011..  Available for
 Petrographic Examination of                         purchase: https://
 Hardened Concrete'', ASTM                           www.astm.org/
 International.                                      Standards/C856.htm.
``Evaluation of Existing      ACI 349.3R-02, June   Available for
 Nuclear Safety Related        2002.                 purchase: https://
 Concrete Structures'',                              www.concrete.org/
 American Concrete Institute.                        store/
                                                     productdetail.aspx?
                                                     ItemID=349302&Forma
                                                     t=DOWNLOAD.
``Guide to the Evaluation     CSA A864-00           Available for
 and Management of Concrete    Reaffirmed 2005.      purchase: https://
 Structures Affected by                              shop.csa.ca/en/
 Alkali-Aggregate                                    canada/concrete/
 Reaction'', CSA Group.                              a864-00-r2005/invt/
                                                     27010172000.
``ASR/DEF Damaged Bent Caps:  Technical Report No.  https://
 Shear Tests and Field         12-8XXIA006, August   library.ctr.utexas.
 Implications'' Texas          2009.                 edu/digitized/
 Department of                                       IACreports/IAC-12-
 Transportation.                                     8XXIA006.pdf.
``Report on the Diagnosis,    FHWA-HIF-09-004,      https://
 Prognosis, and Mitigation     January 2010.         www.fhwa.dot.gov/
 of Alkali-Silica Reaction                           pavement/concrete/
 (ASR) in Transportation                             pubs/hif09004/
 Structures'', Federal                               hif09004.pdf.
 Highway Administration.
NRC Information Notice 2011-  ADAMS Accession No.   https://www.nrc.gov/
 20: Concrete Degradation by   ML112241029,          docs/ML1122/
 Alkali-Silica Reaction, NRC.  November 18, 2011.    ML112241029.pdf.
``Position Paper: In Situ     ADAMS Accession No.   https://www.nrc.gov/
 Monitoring of Alkali-Silica   ML13108A047, April    docs/ML1310/
 Reaction (ASR) Affected       30, 2013.             ML13108A047.pdf.
 Concrete: A Study on Crack
 Indexing and Damage Rating
 Index to Assess the
 Severity of ASR and to
 Monitor ASR Progression'',
 NRC.
------------------------------------------------------------------------
            Referenced Documents Specific to Seabrook Station
------------------------------------------------------------------------
``Seabrook Station: Impact    ADAMS Accession No.   https://www.nrc.gov/
 of Alkali-Silica Reaction     ML12151A397, May      docs/ML1215/
 on Concrete Structures and    2012.                 ML12151A397.pdf.
 Attachments'', MPR
 Associates Inc.
``Seabrook Station Response   ADAMS Accession No.   https://www.nrc.gov/
 to Confirmatory Action        ML13151A328, May 1,   docs/ML1315/
 Letter'', NextEra.            2013.                 ML13151A328.pdf.
Letter from David Wright,     ADAMS Accession No.   https://www.nrc.gov/
 UCS, to NRC Commissioners,    ML13309B606,          docs/ML1330/
 UCS.                          November 4, 2013.     ML13309B606.pdf.
Letter from William M. Dean,  ADAMS Accession No.   https://www.nrc.gov/
 NRC, to David Wright, UCS,    ML13340A405,          docs/ML1334/
 NRC.                          December 6, 2013.     ML13340A405.pdf.
Letter from Robert M.         ADAMS Accession No.   https://www.nrc.gov/
 Taylor, NRC, to Sandra        ML16169A172, July     docs/ML1616/
 Gavutis, C-10, NRC.           6, 2016.              ML16169A172.pdf.
------------------------------------------------------------------------
                     Additional Referenced Documents
------------------------------------------------------------------------
NUREG-1801, ``Generic Aging   December 2010.......  https://www.nrc.gov/
 Lessons Learned Report,''                           reading-rm/doc-
 Revision 2.                                         collections/nuregs/
                                                     staff/sr1801/.

[[Page 65032]]

 
RG 1.160, ``Monitoring the    ADAMS Accession No.   https://www.nrc.gov/
 Effectiveness of              ML113610098, May      docs/ML1136/
 Maintenance at Nuclear        2012.                 ML113610098.pdf.
 Power Plants,'' Revision 3.
``Davis-Besse Nuclear Power   ADAMS Accession No.   https://www.nrc.gov/
 Station_Inspection of        ML15148A489, May      docs/ML1514/
 Apparent Cause Evaluation     28, 2015.             ML15148A489.pdf.
 Efforts for Propagation of
 Laminar Cracking in
 Reinforced Concrete Shield
 Building and Closure of
 Unresolved Item Involving
 Shield Building Laminar
 Cracking Licensing Basis--
 Inspection Report 05000346/
 2014008'', NRC.
------------------------------------------------------------------------


    Dated at Rockville, Maryland, this 19th day of November 2019.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2019-25489 Filed 11-25-19; 8:45 am]
BILLING CODE 7590-01-P