[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]
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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.
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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.
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Submission No. ADAMS accession No. Commenter Affiliation
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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.
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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